Irdf-2002 Dosimetry file - Group data 2002 0 0 0 3.00600E+3 5.96340E+0 0 0 34 10 325 1451 1 0.0 0.0 0 0 0 6 325 1451 2 1.00000E+0 2.00000E+7 0 0 10 2002 325 1451 3 3.00000E+2 0.0 1 0 237 4 325 1451 4 3-Li- 6 LANL EVAL-APR89 G.M.HALE, P.G.YOUNG 325 1451 5 DIST-Feb2004 325 1451 6 ----IRDF-2002 MATERIAL 325 325 1451 7 -----INCIDENT NEUTRON DATA 325 1451 8 ------ENDF-6 FORMAT 325 1451 9 ****************************************************************** 325 1451 10 3-LI - 6 LANL EVAL-APR89 G.M.HALE, P.G.YOUNG 325 1451 11 DIST-SEP91 REV1-JUL91 19910806 325 1451 12 ----ENDF/B-VI MATERIAL 325 REVISION 1 325 1451 13 ****************************************************************** 325 1451 14 3-LI - 6 LANL EVAL-APR89 G.M.HALE, P.G.YOUNG 325 1451 15 DIST-FEB91 910201 325 1451 16 ----ENDF/B-VI MATERIAL 325 325 1451 17 *****************EXTRACT FOR SPECIAL PURPOSE FILE***************** 325 1451 18 DOSIMETRY 325 1451 19 ****************************************************************** 325 1451 20 891221 325 1451 21 ----ENDF/B-VI MATERIAL 325 325 1451 22 325 1451 23 ****************************************************************** 325 1451 24 325 1451 25 ***************************************************************** 325 1451 26 ***************************************************************** 325 1451 27 325 1451 28 MOD1 OF ENDF/B-VI 325 1451 29 325 1451 30 The following revisions were made for MOD1 of ENDF/B-VI: 325 1451 31 325 1451 32 1. MF=1,MT=451 - Comments were added regarding estimated 325 1451 33 (expanded) covariance for the Standards Cross Sections. 325 1451 34 2. MF=3,MT=53 - LF flag and Q-value corrected. 325 1451 35 325 1451 36 ***************************************************************** 325 1451 37 ****************************************************************** 325 1451 38 325 1451 39 ENDF/VI EVALUATION 325 1451 40 G. M. Hale and P. G. Young 325 1451 41 325 1451 42 MAJOR CHANGES FROM VERSION V OF ENDF/B ARE: 325 1451 43 325 1451 44 1. Inclusion of the ENDF/B-VI standard (n,t) cross section 325 1451 45 from the simultaneous standards analysis (ca85) over the 325 1451 46 energy range thermal to 1 MeV. 325 1451 47 2. Replacement of all major cross sections and elastic angular 325 1451 48 distributions at energies between 10^-5 eV and 3 MeV with 325 1451 49 results from the R-matrix analysis performed in conjunction 325 1451 50 with the simultaneous standards analysis. 325 1451 51 3. Revision of the elastic cross sections and angular distri- 325 1451 52 butions at energies between 3 and 20 MeV to match recent 325 1451 53 experimental data, resulting in a general decrease of the 325 1451 54 elastic cross section in this energy range. 325 1451 55 4. Revision of the (n,n')d cross sections to account for 325 1451 56 recent measurements, resulting in a general increase in 325 1451 57 the total (n,n')d cross section that tends to offset the 325 1451 58 decrease in the elastic cross section and maintain about 325 1451 59 the same total cross section as before. 325 1451 60 325 1451 61 325 1451 62 ****************************************************************** 325 1451 63 325 1451 64 STANDARDS COVARIANCES 325 1451 65 325 1451 66 Phase 1 reviewers of the ENDF/B-VI standards cross sections have 325 1451 67 expressed the concern that the uncertainties resulting from the 325 1451 68 combination of R-matrix and simultaneous evaluations might have 325 1451 69 led to uncertainties that are too small. As a result, the 325 1451 70 Standards Subcommittee produced (at the May, 1990 CSEWG meeting) 325 1451 71 a set of expanded covariance estimates for the standard cross 325 1451 72 section reactions. These uncertainties are estimates such that 325 1451 73 if a modern day experiment were performed on a given standard 325 1451 74 cross section using the best techniques, approximately 2/3 of 325 1451 75 the results should fall within these expanded uncertainties. The 325 1451 76 expanded uncertainties for the Li-6(n,t) cross section are given 325 1451 77 in the following table and are compared to values from the 325 1451 78 combined output of the standards covariance analysis: 325 1451 79 325 1451 80 Energy Range Estimated Uncertainty Combined Analysis 325 1451 81 (keV) (percent) (percent) 325 1451 82 325 1451 83 1.0E-08 - 0.1 0.3 0.14 325 1451 84 0.1 - 1.0 0.5 325 1451 85 1.0 - 10. 0.7 0.14 325 1451 86 10. - 50. 0.9 325 1451 87 50. - 90. 1.1 0.25 325 1451 88 90. - 150 1.5 325 1451 89 150 - 450 2.0 0.29 325 1451 90 450 - 650 5.0 325 1451 91 650 - 800 2.0 0.36 325 1451 92 800 - 1000 5.0 325 1451 93 325 1451 94 **************************************************************** 325 1451 95 325 1451 96 325 1451 97 mf=2 --------- resonance parameters ---------------------------- 325 1451 98 325 1451 99 mt=151 effective scattering radius = 2.31175e-13 cm. 325 1451 100 325 1451 101 mf=3 --------- smooth cross sections --------------------------- 325 1451 102 325 1451 103 the 2200 m/s cross sections are as follows: 325 1451 104 mt=1 sigma = 941.6928 barns 325 1451 105 mt=2 sigma = 0.67157 barns 325 1451 106 mt=102 sigma = 0.03850 barns 325 1451 107 mt=105 sigma = 940.9827 barns 325 1451 108 325 1451 109 mt=105 (n,t) cross section 325 1451 110 below 3 mev, values are taken from the r-matrix analysis, 325 1451 111 which includes (n,t) measurements from re78, la78, br77, 325 1451 112 ov74, and ba75. between 3 and 5 mev, the values are 325 1451 113 based on ba75, and at higher energies are taken from the 325 1451 114 evaluation of pe64, extended to 20 mev considering the 325 1451 115 data of ke58. 325 1451 116 325 1451 117 325 1451 118 325 1451 119 mf=33----------cross section covariances------------------------- 325 1451 120 (to be added later) 325 1451 121 325 1451 122 the relative covariances for mt=1,2, and 105 below 4 mev are 325 1451 123 given in file 33. they are based on calculations using the co- 325 1451 124 variances of the r-matrix parameters in first-order error 325 1451 125 propogation. 325 1451 126 mt=1 total 325 1451 127 relative covariances entered as nc-type sub-subsection, 325 1451 128 implying that they are to be constructed from those for 325 1451 129 mt=2 and 105. they are not intended for use at energies 325 1451 130 above 4 mev. 325 1451 131 mt=2,105 elastic and (n,t) 325 1451 132 relative covariances among these two cross sections are 325 1451 133 entered explicitly as ni-type sub-subsections in the 325 1451 134 lb=5 (direct) representation at energies below 4 mev. 325 1451 135 although values for the 3.95-4.05 mev bin are repeated 325 1451 136 in a 4-20 mev bin, the covariances are not intended for 325 1451 137 use at energies above 4 mev. 325 1451 138 325 1451 139 325 1451 140 --------------------- references -------------------------------- 325 1451 141 325 1451 142 ab70 U.Abbondanno, Nuo.Cim. A166,139(1970). 325 1451 143 aj74 f.ajzenberg-selove and t.lauritsen, nucl. phys. a227,55 325 1451 144 (1974). 325 1451 145 ar64 a.h.armstrong, j.gammel, l.rosen, and g.m.frye, nucl. phys. 325 1451 146 52,505 (1964). 325 1451 147 as63 v.j.ashby et al, phys. rev. 129,1771 (1963). 325 1451 148 ba53 m.e.battat and f.l.ribe, phys.rev. 89,80 (1953). 325 1451 149 ba63 r.batchelor and j.h.towle, nucl. phys. 47,385 (1963). 325 1451 150 ba65 r.bass, c.bindhardt, and k.kruger, eandc(e)-57u (1965). 325 1451 151 ba75 c.m.bartle, proc. conf. on nuclear cross sections and 325 1451 152 technology, vol.2,688 (1975), and private communication 325 1451 153 (1976). see also nucl. phys. a330, 1 (1979). 325 1451 154 be75 besotosnyj et al., yk-19, 77 (1975). 325 1451 155 br77 r.e.brown,g.g.ohlsen,r.f.haglund, and n.jarmie, phys. rev. 325 1451 156 16c, 513 (1977). 325 1451 157 ca85 a.d.carlson,w.p.poenitz,g.m.hale, and r.w.peele, nuclear 325 1451 158 data for basic and applied science (santa fe, n.m.), 1429 325 1451 159 (1985). 325 1451 160 co67 j.a.cookson and d.dandy, nucl. phys. a91,273 (1967). 325 1451 161 co82 h.conde,t.andersson,l.nilsson, and c.nordborg, nuclear data 325 1451 162 for science and technology (antwerp, belgium), 447 (1982). 325 1451 163 de73 F.Demanins et al., infn/be-73 (1973). 325 1451 164 dr82 m.drosg,d.m.drake,r.a.hardekopf, and g.m.hale, la-9129-ms 325 1451 165 (1982). 325 1451 166 dr85 m.drosg et al., santa fe conf.1, 145(1985). 325 1451 167 fo71 d.g.foster and d.w.glasgow, phys. rev. c3,576 (1971). 325 1451 168 fr54 g.m.frye, phys. rev. 93,1086 (1954). 325 1451 169 go72 c.a.goulding and p.stoler, eandc(us)-176u,161 (1972). 325 1451 170 ha75 j.a.harvey and n.w.hill, nuclear cross sections and 325 1451 171 technology (washington, d.c.), 244 (1975). 325 1451 172 ha84 g.m.hale, nuclear standard reference data (geel,belgium) 325 1451 173 iaea tecdoc-335, 103 (1984). describes preliminary analysis. 325 1451 174 ho68 j.c.hopkins,d.m.drake, and h.conde, nucl. phys. a107,139 325 1451 175 (1968), and j.c.hopkins, d.m.drake, and h.conde, la-3765 325 1451 176 (1967). 325 1451 177 ho79 h.h.hogue et al., n.s.&e. 69, 22 (1979). 325 1451 178 ju73 e.t.jurney, lasl, private communication (1973). 325 1451 179 ke58 r.d.kern and w.e.kreger, phys. rev. 112, 926 (1958). 325 1451 180 ke79 j.d.kellie,g.p.lamaze, and r.b.schwartz, nuclear cross 325 1451 181 sections for technology (knoxville, tn.), 48 (1979). 325 1451 182 kn77 h.h.knitter,c.budtz-jorgensen,m.mailly, and r.vogt, eur- 325 1451 183 5726e (1977). 325 1451 184 kn79 h.d.knox,r.m.white, and r.o.lane, n.s.&e. 69, 223 (1979). 325 1451 185 kn83 h.h.knitter,c.budtz-jorgensen,d.l.smith, and d.marletta, 325 1451 186 n.s.&e. 83, 229(1983). 325 1451 187 la61 r.o.lane,a.s.langsdorf,j.e.monahan, and a.j.elwyn, ann. 325 1451 188 phys.12, 135 (1961). 325 1451 189 la78 g.p.lamaze,o.a.wasson,r.a.schrack, and a.d.carlson, n.s.&e. 325 1451 190 68, (1978). 325 1451 191 li80 p.w.lisowski et al., la-8342 (1980). 325 1451 192 ma69 d.s.mather and l.f.paine, awre-o-47/69 (1969). 325 1451 193 me65 f.merchez,n.v.sen,v.regis, and r.bouchez, compt. rend. 260, 325 1451 194 3922 (1965). 325 1451 195 ov74 j.c.overley,r.m.sealock, and d.h.ehlers, nucl. phys. a221, 325 1451 196 573 (1974). 325 1451 197 pe64 e.d.pendlebury, awre-o-60/64 (1964). 325 1451 198 pr69 g.presser et al., nuc.phys. a131, 679(1969). 325 1451 199 re78 c.renner,j.a.harvey,n.w.hill,g.l.morgan, and k.pusk, bull. 325 1451 200 am. phys. soc. 23, 526 (1978). 325 1451 201 sa82 e.t.sadowski,h.knox,d.a.resler, and r.o.lane, bap 27,624(c5) 325 1451 202 (1982). 325 1451 203 sm77 a.b.smith,p.guenther,d.havel, and j.f.whalen, anl/ndm-29 325 1451 204 (1977). 325 1451 205 sm82 a.b.smith,p.t.guenther, and j.f.whalen, nucl. phys. a373, 325 1451 206 305 (1982). 325 1451 207 wo62 c.wong,j.d.anderson, and j.w.mcclure, nucl. phys. 33,680 325 1451 208 (1962). 325 1451 209 ****************************************************************** 325 1451 210 325 1451 211 325 1451 212 325 1451 213 325 1451 214 ****************************************************************** 325 1451 215 325 1451 216 ***************** Program LINEAR (VERSION 2002-1) *************** 325 1451 217 For All Data Greater than 1.0000E-10 barns in Absolute Value 325 1451 218 Data Linearized to Within an Accuracy of .100000000 per-cent 325 1451 219 ***************** Program SIGMA1 (VERSION 2002-1) *************** 325 1451 220 Data Doppler Broadened to 300.000000 Kelvin 325 1451 221 for All Data Greater than 1.0000E-10 barns in Absolute Value 325 1451 222 Data Linearized to Within an Accuracy pf .100000000 per-cent 325 1451 223 ***************** Program FIXUP (Version 2002-1) **************** 325 1451 224 Corrected ZA/AWR in All Sections-----------------------------Yes 325 1451 225 Corrected Thresholds-----------------------------------------Yes 325 1451 226 Extended Cross Sections to 20 MeV----------------------------No 325 1451 227 Allow Cross Section Deletion---------------------------------No 325 1451 228 Allow Cross Section Reconstruction---------------------------No 325 1451 229 Make All Cross Sections Non-Negative-------------------------Yes 325 1451 230 Delete Energies Not in Ascending Order-----------------------Yes 325 1451 231 Deleted Duplicate Points-------------------------------------Yes 325 1451 232 Check for Ascending MAT/MF/MT Order--------------------------Yes 325 1451 233 Check for Legal MF/MT Numbers--------------------------------Yes 325 1451 234 Allow Creation of Missing Sections---------------------------No 325 1451 235 Allow Insertion of Energy Points-----------------------------No 325 1451 236 Create Uniform Energy Grid-----------------------------------No 325 1451 237 Delete Section if Cross Section =0 at All Energies-----------Yes 325 1451 238 ***************** Program GROUPIE (VERSION 2002-1) ************** 325 1451 239 Unshielded Group Averages Using 640 Groups 325 1451 240 Weighting Spectrum: Flat (Constant) Spectrum 325 1451 241 1 451 245 2 325 1451 242 2 151 4 2 325 1451 243 3 105 217 1 325 1451 244 33 105 33 1 325 1451 245 325 1 099999 325 0 0 0 3.00600E+3 5.96340E+0 0 0 1 0 325 2151 1 3.006000+3 1.000000+0 0 0 1 0 325 2151 2 1.000000-5 1.000000+5 0 0 0 0 325 2151 3 1.000000+0 2.311750-1 0 0 0 0 325 2151 4 325 2 099999 325 0 0 0 3.00600E+3 5.96340E+0 0 0 0 0 325 3105 1 4.78380E+6 4.78380E+6 0 0 1 641 325 3105 2 641 1 325 3105 3 .000100000 14794.9461 .000105000 14449.9534 .000110000 14122.2061 325 3105 4 .000115000 13820.1942 .000120000 13467.1458 .000127500 13075.4530 325 3105 5 .000135000 12717.5721 .000142500 12388.6847 .000150000 12032.7367 325 3105 6 .000160000 11661.9312 .000170000 11324.5439 .000180000 11014.8459 325 3105 7 .000190000 10726.6978 .000200000 10463.6234 .000210000 10216.1135 325 3105 8 .000220000 9986.21324 .000230000 9772.62292 .000240000 9521.91859 325 3105 9 .000255000 9246.40272 .000270000 9033.79728 .000280000 8797.12696 325 3105 10 .000300000 8508.50325 .000320000 8247.16668 .000340000 8007.82467 325 3105 11 .000360000 7787.37574 .000380000 7586.13528 .000400000 7376.33068 325 3105 12 .000425000 7161.64277 .000450000 6965.50583 .000475000 6785.66231 325 3105 13 .000500000 6616.52235 .000525000 6461.94757 .000550000 6315.93048 325 3105 14 .000575000 6180.13284 .000600000 6040.94642 .000630000 5897.75751 325 3105 15 .000660000 5766.73697 .000690000 5641.05202 .000720000 5507.32050 325 3105 16 .000760000 5363.93249 .000800000 5230.80234 .000840000 5108.78635 325 3105 17 .000880000 4992.81623 .000920000 4886.50890 .000960000 4784.70637 325 3105 18 .001000000 4679.19964 .001050000 4569.04414 .001100000 4465.76191 325 3105 19 .001150000 4370.59678 .001200000 4258.33842 .001275000 4134.96155 325 3105 20 .001350000 4022.16216 .001425000 3916.88925 .001500000 3805.04153 325 3105 21 .001600000 3688.10030 .001700000 3581.34573 .001800000 3482.60854 325 3105 22 .001900000 3392.40441 .002000000 3308.77765 .002100000 3230.31157 325 3105 23 .002200000 3158.44480 .002300000 3089.72099 .002400000 3011.30697 325 3105 24 .002550000 2924.17510 .002700000 2856.02126 .002800000 2781.96182 325 3105 25 .003000000 2690.80549 .003200000 2607.85391 .003400000 2531.98785 325 3105 26 .003600000 2462.79638 .003800000 2398.91608 .004000000 2332.33692 325 3105 27 .004250000 2264.72689 .004500000 2203.02081 .004750000 2145.35559 325 3105 28 .005000000 2092.44576 .005250000 2043.40654 .005500000 1997.05482 325 3105 29 .005750000 1954.57647 .006000000 1909.96714 .006300000 1865.38391 325 3105 30 .006600000 1823.18871 .006900000 1784.04467 .007200000 1741.48065 325 3105 31 .007600000 1695.96373 .008000000 1654.34447 .008400000 1615.27928 325 3105 32 .008800000 1578.91854 .009200000 1545.13798 .009600000 1513.00017 325 3105 33 .010000000 1479.73644 .010500000 1444.57603 .011000000 1412.42313 325 3105 34 .011500000 1381.75051 .012000000 1346.59615 .012750000 1307.68584 325 3105 35 .013500000 1271.64181 .014250000 1238.56317 .015000000 1203.27730 325 3105 36 .016000001 1166.23862 .017000001 1132.28052 .017999999 1101.28202 325 3105 37 .018999999 1072.79317 .020000000 1046.06875 .021000000 1021.61496 325 3105 38 .022000000 998.572690 .023000000 977.099923 .024000000 951.867476 325 3105 39 .025500000 924.650334 .027000001 903.085714 .028000001 879.664218 325 3105 40 .029999999 850.841313 .032000002 824.584971 .034000002 800.612528 325 3105 41 .035999998 778.732778 .037999999 758.549586 .039999999 737.487875 325 3105 42 .042500000 716.110259 .045000002 696.571444 .047499999 678.366928 325 3105 43 .050000001 661.623345 .052499998 646.130877 .055000000 631.464013 325 3105 44 .057500001 618.035173 .059999999 603.916290 .063000001 589.802419 325 3105 45 .066000000 576.497152 .068999998 564.095301 .071999997 550.648650 325 3105 46 .075999998 536.246480 .079999998 523.089245 .083999999 510.719324 325 3105 47 .088000000 499.211911 .092000000 488.553833 .096000001 478.374129 325 3105 48 .100000001 467.864268 .104999997 456.740352 .109999999 446.578283 325 3105 49 .115000002 436.871661 .119999997 425.753899 .127499998 413.473559 325 3105 50 .135000005 402.085828 .142499998 391.626836 .150000006 380.471804 325 3105 51 .159999996 368.744950 .170000002 358.015375 .180000007 348.198886 325 3105 52 .189999998 339.184378 .200000003 330.720268 .209999993 322.982244 325 3105 53 .219999999 315.682189 .230000004 308.885854 .239999995 301.048031 325 3105 54 .254999995 292.298232 .270000011 285.625887 .280000001 278.184188 325 3105 55 .300000012 269.039086 .319999993 260.740769 .340000004 253.189079 325 3105 56 .360000014 246.269336 .379999995 239.820982 .400000006 233.210767 325 3105 57 .425000012 226.472440 .449999988 220.243253 .474999994 214.511408 325 3105 58 .500000000 209.256293 .524999976 204.282890 .550000012 199.729453 325 3105 59 .574999988 195.403926 .600000024 190.991063 .629999995 186.517846 325 3105 60 .660000026 182.279486 .689999998 178.400407 .720000029 174.092531 325 3105 61 .759999990 169.580921 .800000012 165.402550 .839999974 161.482057 325 3105 62 .879999995 157.896838 .920000017 154.457639 .959999979 151.313264 325 3105 63 1.00000000 147.924471 1.04999995 144.457738 1.10000002 141.215388 325 3105 64 1.14999998 138.163257 1.20000005 134.661002 1.27499998 130.732247 325 3105 65 1.35000002 127.134297 1.42499995 123.849352 1.50000000 120.293925 325 3105 66 1.60000002 116.585459 1.70000005 113.209752 1.79999995 110.125223 325 3105 67 1.89999998 107.244386 2.00000000 104.595170 2.09999990 102.145940 325 3105 68 2.20000005 99.8337043 2.29999995 97.7171215 2.40000010 95.2040034 325 3105 69 2.54999995 92.4334553 2.70000005 90.3148276 2.79999995 87.9381155 325 3105 70 3.00000000 85.0523601 3.20000005 82.4380221 3.40000010 80.0560785 325 3105 71 3.59999990 77.8616733 3.79999995 75.8730582 4.00000000 73.7205775 325 3105 72 4.25000000 71.6078328 4.50000000 69.6686317 4.75000000 67.8455059 325 3105 73 5.00000000 66.1838277 5.25000000 64.5853545 5.50000000 63.1383990 325 3105 74 5.75000000 61.7747034 6.00000000 60.3789510 6.30000019 58.9664576 325 3105 75 6.59999990 57.6264409 6.90000010 56.4004006 7.19999981 55.0410801 325 3105 76 7.59999990 53.6130860 8.00000000 52.2937668 8.39999962 51.0497077 325 3105 77 8.80000019 49.9143943 9.19999981 48.8331393 9.60000038 47.8550339 325 3105 78 10.0000000 46.7674076 10.5000000 45.6690583 11.0000000 44.6357610 325 3105 79 11.5000000 43.7022554 12.0000000 42.5635629 12.7500000 41.3292052 325 3105 80 13.5000000 40.1934306 14.2500000 39.1440984 15.0000000 38.0253790 325 3105 81 16.0000000 36.8529448 17.0000000 35.7847221 18.0000000 34.8158482 325 3105 82 19.0000000 33.8933511 20.0000000 33.0523746 21.0000000 32.2802396 325 3105 83 22.0000000 31.5498964 23.0000000 30.8900598 24.0000000 30.0863941 325 3105 84 25.5000000 29.2029000 27.0000000 28.5498764 28.0000000 27.7953130 325 3105 85 30.0000000 26.8770847 32.0000000 26.0473656 34.0000000 25.2913699 325 3105 86 36.0000000 24.6097325 38.0000000 23.9570474 40.0000000 23.2925112 325 3105 87 42.5000000 22.6200978 45.0000000 21.9989139 47.5000000 21.4210590 325 3105 88 50.0000000 20.8953058 52.5000000 20.3989799 55.0000000 19.9413886 325 3105 89 57.5000000 19.5133607 60.0000000 19.0727461 63.0000000 18.6328468 325 3105 90 66.0000000 18.2080695 69.0000000 17.8128281 72.0000000 17.3948470 325 3105 91 76.0000000 16.9347672 80.0000000 16.5172078 84.0000000 16.1339473 325 3105 92 88.0000000 15.7710512 92.0000000 15.4221555 96.0000000 15.1053270 325 3105 93 100.000000 14.7700328 105.000000 14.4288661 110.000000 14.0965950 325 3105 94 115.000000 13.8008879 120.000000 13.4398837 127.500000 13.0466643 325 3105 95 135.000000 12.6979150 142.500000 12.3611838 150.000000 12.0094109 325 3105 96 160.000000 11.6403331 170.000000 11.2996687 180.000000 10.9872033 325 3105 97 190.000000 10.7071240 200.000000 10.4367810 210.000000 10.1921719 325 3105 98 220.000000 9.96131944 230.000000 9.74939037 240.000000 9.49846206 325 3105 99 255.000000 9.22265603 270.000000 9.01502091 280.000000 8.77535127 325 3105 100 300.000000 8.48444616 320.000000 8.22571754 340.000000 7.98405981 325 3105 101 360.000000 7.76663440 380.000000 7.56467016 400.000000 7.35376806 325 3105 102 425.000000 7.13932704 450.000000 6.94443603 475.000000 6.76435003 325 3105 103 500.000000 6.59661622 525.000000 6.44315602 550.000000 6.29397110 325 3105 104 575.000000 6.15893198 600.000000 6.02324074 630.000000 5.87760877 325 3105 105 660.000000 5.74686517 690.000000 5.62296263 720.000000 5.48674802 325 3105 106 760.000000 5.34606717 800.000000 5.21088095 840.000000 5.08749818 325 3105 107 880.000000 4.97416821 920.000000 4.86865092 960.000000 4.76557776 325 3105 108 1000.00000 4.66214100 1050.00000 4.55068283 1100.00000 4.44663325 325 3105 109 1150.00000 4.35342467 1200.00000 4.23972217 1275.00000 4.11709895 325 3105 110 1350.00000 4.00425444 1425.00000 3.89861606 1500.00000 3.78734478 325 3105 111 1600.00000 3.67248621 1700.00000 3.56544145 1800.00000 3.46853432 325 3105 112 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4.541172-7 4.824173-7 5.252231-6 6.679399-6 5.083272-6 32533105 24 1.396229-6 2.897295-7 2.358371-7 2.244136-6 2.418936-6 7.837799-7 32533105 25 7.614119-7 9.288824-6 8.271248-6 2.397961-6 9.349064-7 9.954863-9 32533105 26 2.405782-6 3.039443-6 1.254285-6 1.052807-6 1.105814-5 4.339311-6 32533105 27 2.488444-6 9.852444-7 1.377236-6 1.989021-6 1.457171-6 9.981584-7 32533105 28 4.886551-6 3.473981-6 1.653619-6 9.322945-7 7.297434-7 6.455541-7 32533105 29 7.013815-7 8.443410-6 5.612672-6 2.134349-6 3.386685-7 2.943248-8 32533105 30 7.915241-7 9.299034-6 8.280879-6 5.098584-6 1.338568-6 1.841448-6 32533105 31 1.270541-5 1.262244-5 7.390554-6 4.714562-6 1.835564-5 1.767288-5 32533105 32 9.582569-6 3.027060-5 1.780456-5 3.434466-5 32533105 33 32533 099999 325 0 0 0 0 0 0 0 5.00000E+3 1.07191E+1 -1 0 34 0 500 1451 1 0.0 0.0 0 0 0 6 500 1451 2 1.00000E+0 2.00000E+7 0 0 10 2002 500 1451 3 3.00000E+2 0.0 1 0 75 5 500 1451 4 5-B - 0 NDS IAEA-JAN04 500 1451 5 DIST-Feb2004 500 1451 6 ----IRDF-2002 500 1451 7 -----INCIDENT NEUTRON DATA 500 1451 8 ------ENDF-6 FORMAT 500 1451 9 ***************************************************************** 500 1451 10 *****************EXTRACT FOR SPECIAL PURPOSE FILE***************** 500 1451 11 DOSIMETRY Assembled at NDS from the ENDF/B-VI evaluations for 500 1451 12 5-B -10 and 5-B -11. The files were processed 500 1451 13 through the 2002 pre-processing codes LINEAR,RECENT, 500 1451 14 SIGMA1 and FIXUP to produce pointwise files prior to 500 1451 15 input to MIXER. 500 1451 16 500 1451 17 MIXER INPUT FILE 500 1451 18 500 1451 19 Boron DENSITY 2.34 G/CC. CONSTITUENTS: 10-19.8 11-80.2 500 1451 20 FIXUP.out 500 1451 21 MIXER.OUT 500 1451 22 5000 500 3 1 (ID FOR COMBINED REACTION) 500 1451 23 5010 1 0.429086 500 1451 24 5011 1 1.910914 500 1451 25 (BLANK LINE TERMINATES LIST) 500 1451 26 ****************************************************************** 500 1451 27 500 1451 28 5-B - 10 LANL EVAL-NOV89 G.M.HALE, P.G.YOUNG 500 1451 29 DIST-SEP91 REV1-JUL91 19910806 500 1451 30 ----ENDF/B-VI MATERIAL 525 REVISION 1 500 1451 31 ***************************************************************** 500 1451 32 5-B - 11 LANL EVAL-MAY89 P.G.YOUNG 500 1451 33 DIST-SEP 1 REV1- 20010926 500 1451 34 ----ENDF/B-VI MATERIAL 528 500 1451 35 ***************** Program MIXER (VERSION 2002-1) **************** 500 1451 36 Boron DENSITY 2.34 G/CC. CONSTITUENTS: 10-19.8 500 1451 37 11-80.2 500 1451 38 ---------------------------------------- 500 1451 39 Composition 500 1451 40 ---------------------------------------- 500 1451 41 Isotope MF MT Atom-Fract Grams/cc 500 1451 42 ---------------------------------------- 500 1451 43 5010 3 1 .198003533 .429086000 500 1451 44 5011 3 1 .801996467 1.91091400 500 1451 45 ---------------------------------------- 500 1451 46 5000 3 1 1.00000000 2.34000000 500 1451 47 ---------------------------------------- 500 1451 48 Composition 500 1451 49 ---------------------------------------- 500 1451 50 Isotope MF MT Atom-Fract Grams/cc 500 1451 51 ---------------------------------------- 500 1451 52 5010 3 101 .198003533 .429086000 500 1451 53 5011 3 101 .801996467 1.91091400 500 1451 54 ---------------------------------------- 500 1451 55 5000 3 101 1.00000000 2.34000000 500 1451 56 ---------------------------------------- 500 1451 57 Composition 500 1451 58 ---------------------------------------- 500 1451 59 Isotope MF MT Atom-Fract Grams/cc 500 1451 60 ---------------------------------------- 500 1451 61 5010 3 102 .198003533 .429086000 500 1451 62 5011 3 102 .801996467 1.91091400 500 1451 63 ---------------------------------------- 500 1451 64 5000 3 102 1.00000000 2.34000000 500 1451 65 ---------------------------------------- 500 1451 66 Composition 500 1451 67 ---------------------------------------- 500 1451 68 Isotope MF MT Atom-Fract Grams/cc 500 1451 69 ---------------------------------------- 500 1451 70 5010 3 107 .198003533 .429086000 500 1451 71 5011 3 107 .801996467 1.91091400 500 1451 72 ---------------------------------------- 500 1451 73 5000 3 107 1.00000000 2.34000000 500 1451 74 ---------------------------------------- 500 1451 75 ***************************************************************** 500 1451 76 ***************** Program GROUPIE (VERSION 2002-1) ************** 500 1451 77 Unshielded Group Averages Using 640 Groups 500 1451 78 Weighting Spectrum: 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5750.00000 1.53833792 6000.00000 1.50282955 6300.00000 1.46794607 500 3107 120 6600.00000 1.43396741 6900.00000 1.40275447 7200.00000 1.36870201 500 3107 121 7600.00000 1.33180505 8000.00000 1.29798250 8400.00000 1.26704130 500 3107 122 8800.00000 1.23771594 9200.00000 1.21159197 9600.00000 1.18599691 500 3107 123 10000.0000 1.15875085 10500.0000 1.13132325 11000.0000 1.10606930 500 3107 124 11500.0000 1.08140960 12000.0000 1.05331716 12750.0000 1.02219711 500 3107 125 13500.0000 .993621235 14250.0000 .967522791 15000.0000 .939970163 500 3107 126 16000.0000 .910461974 17000.0000 .883955117 18000.0000 .859670891 500 3107 127 19000.0000 .837026687 20000.0000 .816403814 21000.0000 .796992000 500 3107 128 22000.0000 .778972323 23000.0000 .762303468 24000.0000 .742959136 500 3107 129 25500.0000 .721879013 27000.0000 .705630964 28000.0000 .687423555 500 3107 130 30000.0000 .665434100 32000.0000 .645633106 34000.0000 .627841284 500 3107 131 36000.0000 .611336387 38000.0000 .595948890 40000.0000 .580403134 500 3107 132 42500.0000 .563923221 45000.0000 .549508500 47500.0000 .536231320 500 3107 133 50000.0000 .524163021 52500.0000 .512878694 55000.0000 .502309402 500 3107 134 57500.0000 .492977710 60000.0000 .482835278 63000.0000 .472562256 500 3107 135 66000.0000 .463570968 69000.0000 .454898608 72000.0000 .445548458 500 3107 136 76000.0000 .435809400 80000.0000 .426517455 84000.0000 .418246514 500 3107 137 88000.0000 .410927914 92000.0000 .403651642 96000.0000 .396639053 500 3107 138 100000.000 .389433793 105000.000 .382003081 110000.000 .375165493 500 3107 139 115000.000 .368436319 120000.000 .360702442 127500.000 .351572543 500 3107 140 135000.000 .343241912 142500.000 .335345435 150000.000 .326329138 500 3107 141 160000.000 .316537289 170000.000 .306783115 180000.000 .297358661 500 3107 142 190000.000 .288201774 200000.000 .279002174 210000.000 .270048673 500 3107 143 220000.000 .261393856 230000.000 .253065477 240000.000 .242985767 500 3107 144 255000.000 .231576475 270000.000 .222530053 280000.000 .212176815 500 3107 145 300000.000 .199518728 320000.000 .188801004 340000.000 .179927189 500 3107 146 360000.000 .173514984 380000.000 .169522719 400000.000 .167732745 500 3107 147 425000.000 .167367298 450000.000 .166376264 475000.000 .161645611 500 3107 148 500000.000 .152153769 525000.000 .139774643 550000.000 .126881860 500 3107 149 575000.000 .114767061 600000.000 .103279598 630000.000 .092720183 500 3107 150 660000.000 .084094035 690000.000 .076876211 720000.000 .070034625 500 3107 151 760000.000 .063669802 800000.000 .058497154 840000.000 .054252351 500 3107 152 880000.000 .050730286 920000.000 .047795793 960000.000 .045316291 500 3107 153 1000000.00 .042654072 1100000.00 .041979779 1200000.00 .043619696 500 3107 154 1300000.00 .047472097 1400000.00 .053984775 1500000.00 .065055907 500 3107 155 1600000.00 .084445845 1700000.00 .100933533 1800000.00 .102736791 500 3107 156 1900000.00 .095794408 2000000.00 .084696977 2100000.00 .072437129 500 3107 157 2200000.00 .063492741 2300000.00 .058306439 2400000.00 .056201749 500 3107 158 2500000.00 .060131741 2600000.00 .066513985 2700000.00 .074986446 500 3107 159 2800000.00 .075947111 2900000.00 .068738566 3000000.00 .062584334 500 3107 160 3100000.00 .057801534 3200000.00 .055856603 3300000.00 .057232663 500 3107 161 3400000.00 .060268960 3500000.00 .063636930 3600000.00 .067074281 500 3107 162 3700000.00 .070528099 3800000.00 .071525819 3900000.00 .070040466 500 3107 163 4000000.00 .065564482 4100000.00 .058097630 4200000.00 .052996540 500 3107 164 4300000.00 .050076819 4400000.00 .046287789 4500000.00 .041843234 500 3107 165 4600000.00 .038190148 4700000.00 .035338142 4800000.00 .032754060 500 3107 166 4900000.00 .030428289 5000000.00 .028475508 5100000.00 .026895717 500 3107 167 5200000.00 .025327931 5300000.00 .023772152 5400000.00 .022631485 500 3107 168 5500000.00 .022232566 5600000.00 .022325351 5700000.00 .022583208 500 3107 169 5800000.00 .022802486 5900000.00 .022972725 6000000.00 .023142792 500 3107 170 6100000.00 .023312859 6200000.00 .023406626 6300000.00 .023424091 500 3107 171 6400000.00 .023372579 6500000.00 .023252092 6600000.00 .023078708 500 3107 172 6700000.00 .022852427 6800000.00 .022521864 6900000.00 .022087019 500 3107 173 7000000.00 .021559718 7100000.00 .020939960 7200000.00 .020292928 500 3107 174 7300000.00 .019618622 7400000.00 .018951384 7500000.00 .018291300 500 3107 175 7600000.00 .017662035 7700000.00 .017074572 7800000.00 .016584340 500 3107 176 7900000.00 .016216828 8000000.00 .015872377 8100000.00 .015536219 500 3107 177 8200000.00 .015266814 8300000.00 .015068258 8400000.00 .014936052 500 3107 178 8500000.00 .014871101 8600000.00 .014867289 8700000.00 .014920678 500 3107 179 8800000.00 .015022340 8900000.00 .015167674 9000000.00 .015351632 500 3107 180 9100000.00 .015610118 9200000.00 .015963544 9300000.00 .016408857 500 3107 181 9400000.00 .016932279 9500000.00 .017524296 9600000.00 .018153582 500 3107 182 9700000.00 .018815146 9800000.00 .019556606 9900000.00 .020397550 500 3107 183 10000000.0 .021312872 10100000.0 .022308480 10200000.0 .023350896 500 3107 184 10300000.0 .024439042 10400000.0 .025543706 10500000.0 .026643313 500 3107 185 10600000.0 .027694449 10700000.0 .028673826 10800000.0 .029506016 500 3107 186 10900000.0 .030191021 11000000.0 .030763852 11100000.0 .031224509 500 3107 187 11200000.0 .031631834 11300000.0 .031985825 11400000.0 .032291818 500 3107 188 11500000.0 .032568346 11600000.0 .032822288 11700000.0 .033035107 500 3107 189 11800000.0 .033211108 11900000.0 .033350291 12000000.0 .033508437 500 3107 190 12100000.0 .033685547 12200000.0 .033862657 12300000.0 .034039767 500 3107 191 12400000.0 .034204313 12500000.0 .034369036 12600000.0 .034518832 500 3107 192 12700000.0 .034640959 12800000.0 .034721523 12900000.0 .034760524 500 3107 193 13000000.0 .034775931 13100000.0 .034767745 13200000.0 .034759560 500 3107 194 13300000.0 .034751374 13400000.0 .034726948 13500000.0 .034684555 500 3107 195 13600000.0 .034640435 13700000.0 .034596533 13800000.0 .034538519 500 3107 196 13900000.0 .034465304 14000000.0 .034379336 14100000.0 .034280613 500 3107 197 14200000.0 .034181891 14300000.0 .034083168 14400000.0 .033984445 500 3107 198 14500000.0 .033868518 14600000.0 .033725226 14700000.0 .033571775 500 3107 199 14800000.0 .033418325 14900000.0 .033264874 15000000.0 .033090006 500 3107 200 15100000.0 .032893719 15200000.0 .032676905 15300000.0 .032439563 500 3107 201 15400000.0 .032202220 15500000.0 .031962973 15600000.0 .031700025 500 3107 202 15700000.0 .031415284 15800000.0 .031130543 15900000.0 .030845802 500 3107 203 16000000.0 .030542294 16100000.0 .030220019 16200000.0 .029897745 500 3107 204 16300000.0 .029575470 16400000.0 .029241070 16500000.0 .028895700 500 3107 205 16600000.0 .028551489 16700000.0 .028207277 16800000.0 .027863066 500 3107 206 16900000.0 .027518854 17000000.0 .027174643 17100000.0 .026830432 500 3107 207 17200000.0 .026486220 17300000.0 .026142009 17400000.0 .025797797 500 3107 208 17500000.0 .025453586 17600000.0 .025109374 17700000.0 .024765163 500 3107 209 17800000.0 .024430447 17900000.0 .024105228 18000000.0 .023780009 500 3107 210 18100000.0 .023454789 18200000.0 .023129570 18300000.0 .022804351 500 3107 211 18400000.0 .022488956 18500000.0 .022183385 18600000.0 .021877815 500 3107 212 18700000.0 .021572244 18800000.0 .021266674 18900000.0 .020961103 500 3107 213 19000000.0 .020663773 19100000.0 .020374684 19200000.0 .020085594 500 3107 214 19300000.0 .019796505 19400000.0 .019507415 19500000.0 .019218326 500 3107 215 19600000.0 .018935365 19700000.0 .018658533 19800000.0 .018381700 500 3107 216 19900000.0 .018104868 20000000.0 0.0 500 3107 217 500 3 099999 500 0 0 0 0 0 0 0 5.01000E+3 9.92692E+0 0 0 34 10 525 1451 1 0.0 0.0 0 0 0 6 525 1451 2 1.00000E+0 2.00000E+7 0 0 10 2002 525 1451 3 3.00000E+2 0.0 1 0 256 6 525 1451 4 5-B - 10 LANL EVAL-NOV89 G.M.HALE, P.G.YOUNG 525 1451 5 DIST-Feb2004 525 1451 6 ----IRDF-2002 MATERIAL 525 525 1451 7 -----INCIDENT NEUTRON DATA 525 1451 8 ------ENDF-6 FORMAT 525 1451 9 ****************************************************************** 525 1451 10 5-B - 10 LANL EVAL-NOV89 G.M.HALE, P.G.YOUNG 525 1451 11 DIST-SEP91 REV1-JUL91 19910806 525 1451 12 ----ENDF/B-VI MATERIAL 525 REVISION 1 525 1451 13 ****************************************************************** 525 1451 14 5-B - 10 LANL EVAL-NOV89 G.M.HALE, P.G.YOUNG 525 1451 15 DIST-FEB91 910201 525 1451 16 ----IRDF-90 MATERIAL 525 525 1451 17 525 1451 18 *****************EXTRACT FOR SPECIAL PURPOSE FILE***************** 525 1451 19 DOSIMETRY 525 1451 20 ****************************************************************** 525 1451 21 525 1451 22 MOD1 OF ENDF/B-VI 525 1451 23 525 1451 24 The following revisions were made for MOD1 of ENDF/B-VI: 525 1451 25 525 1451 26 1. MF=1,MT=451 - Comments were added regarding estimated 525 1451 27 (expanded) covariance for the Standards Cross Sections. 525 1451 28 2. MF=3,MT=55,57,62,64,65,68,70,71,73,74,76-81,83,84 - LR flags 525 1451 29 and Q-values corrected. 525 1451 30 525 1451 31 ***************************************************************** 525 1451 32 ****************************************************************** 525 1451 33 525 1451 34 ENDF/VI EVALUATION 525 1451 35 G. M. Hale and P. G. Young 525 1451 36 525 1451 37 MAJOR CHANGES FROM VERSION V OF ENDF/B ARE: 525 1451 38 525 1451 39 1. Inclusion of the ENDF/B-VI standard (n,alpha) and (n,alpha1) 525 1451 40 results from the simultaneous standards analysis (Ca85) over 525 1451 41 the standard energy range thermal to 100 keV. 525 1451 42 2. Replacement of all major cross sections and elastic angular 525 1451 43 distributions from 10-5 eV to 1 MeV with results from the 525 1451 44 R-matrix analysis performed in conjunction with the 525 1451 45 simultaneous standards analysis. 525 1451 46 3. Replaced the total cross section 1-20 MeV with results 525 1451 47 from a covariance analysis of available data. 525 1451 48 4. Revised elastic and inelastic cross sections for low-lying 525 1451 49 levels incorporating new elastic, inelastic, and (n,xgamma) 525 1451 50 experimental data. We attempted to better reconcile the 525 1451 51 inelastic and gamma ray data. 525 1451 52 5. Refit all elastic angular distributions from 1-20 MeV with 525 1451 53 Legendre expansions and incorporated results from new 525 1451 54 measurements. 525 1451 55 6. Fit inelastic neutron angular distributions for first 5 525 1451 56 excited states of B10 with Legendre expansions. 525 1451 57 7. Incorporated new (n,t2alpha) cross section data into MT113 525 1451 58 and adjusted (n,alpha) cross sections above standard region 525 1451 59 for better consistency with data as well as other cross 525 1451 60 sections (esp. total and elastic) determined by data. 525 1451 61 525 1451 62 *****Note that covariance data will be added at a later date. 525 1451 63 525 1451 64 ****************************************************************** 525 1451 65 525 1451 66 STANDARDS COVARIANCES 525 1451 67 525 1451 68 Phase 1 reviewers of the ENDF/B-VI standards cross sections have 525 1451 69 expressed the concern that the uncertainties resulting from the 525 1451 70 combination of R-matrix and simultaneous evaluations might have 525 1451 71 led to uncertainties that are too small. As a result, the 525 1451 72 Standards Subcommittee produced (at the May, 1990 CSEWG meeting) 525 1451 73 a set of expanded covariance estimates for the standard cross 525 1451 74 section reactions. These uncertainties are estimates such that 525 1451 75 if a modern day experiment were performed on a given standard 525 1451 76 cross section using the best techniques, approximately 2/3 of 525 1451 77 the results should fall within these expanded uncertainties. The 525 1451 78 expanded uncertainties for the B-10(n,alpha0) and B-10(n,alpha1) 525 1451 79 cross sections are given in the following tables and are 525 1451 80 compared to values from the combined output of the standards 525 1451 81 covariance analysis: 525 1451 82 525 1451 83 525 1451 84 B-10(n,alpha0) Cross Section 525 1451 85 525 1451 86 Energy Range Estimated Uncertainty Combined Analysis 525 1451 87 (keV) (percent) (percent) 525 1451 88 525 1451 89 1.0E-08 - 0.1 0.5 0.21 525 1451 90 0.1 - 5.0 1.5 525 1451 91 5.0 - 30. 3.0 0.38 525 1451 92 30. - 90. 5.0 525 1451 93 90. - 150 10.0 0.86 525 1451 94 150 - 200 12.0 525 1451 95 200 - 250 15.0 0.79 525 1451 96 525 1451 97 525 1451 98 525 1451 99 B-10(n,alpha1) Cross Section 525 1451 100 525 1451 101 Energy Range Estimated Uncertainty Combined Analysis 525 1451 102 (keV) (percent) (percent) 525 1451 103 525 1451 104 1.0E-08 - 0.1 0.2 0.16 525 1451 105 0.1 - 5.0 0.4 525 1451 106 5.0 - 30. 0.6 0.20 525 1451 107 30. - 90. 1.0 525 1451 108 90. - 150 1.5 0.48 525 1451 109 150 - 200 2.0 525 1451 110 200 - 250 2.5 0.62 525 1451 111 525 1451 112 **************************************************************** 525 1451 113 525 1451 114 mf=2 --------- resonance parameters ---------------------------- 525 1451 115 525 1451 116 mt=151 effective scattering radius = 4.129038-13 cm 525 1451 117 525 1451 118 mf=3 --------- smooth cross sections --------------------------- 525 1451 119 525 1451 120 the 2200 m/s cross sections are as follows, 525 1451 121 mt=1 sigma = 3842.146 barns 525 1451 122 mt=2 sigma = 2.142435 barns 525 1451 123 mt=102 sigma = 0.5 barns 525 1451 124 mt=103 sigma = 0.000566 barns 525 1451 125 mt=107 sigma = 3839.496 barns 525 1451 126 mt=113 sigma = 0.0069993 barns 525 1451 127 mt=600 sigma = 0.000566 barns 525 1451 128 mt=800 sigma = 241.2677 barns 525 1451 129 mt=801 sigma = 3598.228 barns 525 1451 130 525 1451 131 mt=1 total cross section 525 1451 132 0 to 1 mev, calculated from r-matrix parameters obtained 525 1451 133 from simultaneous standards analysis (ca85) used to 525 1451 134 obtain the endf/b-vi standard cross sections. 525 1451 135 1 to 20 mev, covariance analysis of measurements of di67, 525 1451 136 ts62,fo61,co52,au79, and co54, constrained to match 525 1451 137 r-matrix fit at 1 mev. glucs covariance analysis code 525 1451 138 (he80) was used in the calculations. 525 1451 139 525 1451 140 525 1451 141 mt=107 (n,alpha) cross section 525 1451 142 0 to 20 mev, sum of mt=800,801. 525 1451 143 525 1451 144 mt=800 (n,alpha0) cross section 525 1451 145 0 to 1 mev, calculated from the r-matrix parameters 525 1451 146 described for mt=1. experimental (n,alpha0) data input 525 1451 147 to the fit were those of ma68 and da61. in addition, the 525 1451 148 angular distributions of va72 for the inverse reaction 525 1451 149 were included in the analysis. 525 1451 150 1 to 20 mev, based on da61 measurements, with smooth extra- 525 1451 151 polation from 8 to 20 mev using 14-MeV data of an69. The 525 1451 152 da61 data above approximately 2 mev were renormalized 525 1451 153 by a factor of approximately 1.4. Note that some of the 525 1451 154 structure seen in da61 was expanded to give consistent 525 1451 155 nonelastic, elastic, and total cross sections when 525 1451 156 compared with experimental data. 525 1451 157 525 1451 158 mt=801 (n,alpha1) cross section 525 1451 159 0 to 1 mev, calculated from the r-matrix parameters 525 1451 160 described for mt=1. experimental (n,alpha1) data in- 525 1451 161 cluded in the fit are those of sc76. in addition, the 525 1451 162 absolute differential cross-section measurements of 525 1451 163 se76 were included in the analysis. 525 1451 164 1 to 20 mev, smooth curve through measurements of da61 and 525 1451 165 ne70, with smooth extrapolation from 15 to 20 mev. the 525 1451 166 da61 data above approximately 2 mev were renormalized 525 1451 167 by a factor of approximately 1.4. Note that some of the 525 1451 168 structure seen in da61 was expanded to give consistent 525 1451 169 nonelastic, elastic, and total cross sections when 525 1451 170 compared with experimental data. 525 1451 171 525 1451 172 525 1451 173 525 1451 174 ----------------------- references ----------------------------- 525 1451 175 525 1451 176 aj75 f. ajzenberg-selove, nucl. phys. a248,6 (1975) 525 1451 177 aj88 f. ajzenberg-selove, nucl. phys. a490,1 (1988) 525 1451 178 an69 B. Antolkovic, Nuc.Phys.A139, 10 (1969). 525 1451 179 as70 a. asami and m.c. moxon, j.nucl.energy 24,85 (1970) 525 1451 180 au79 g.auchampaugh et al., nucl.sci.eng.69,30(1979) 525 1451 181 ba60 r.bardes and g.e. owen, phys.rev.120,1369 (1960) 525 1451 182 be56 r.l. becker and h.h. barschall, phys.rev.102,1384 (1956) 525 1451 183 bo51 c.k.bockelman et al., phys.rev. 84,69 (1951) 525 1451 184 bo69 d.bogart and l.l.nichols, nucl.phys.a125,463 (1969) 525 1451 185 ca85 a.carlson et al., nuc.data for basic & applied science, 525 1451 186 santa fe, nm (1985) p.1429. 525 1451 187 co52 j.h.coon et al., phys.rev. 88,562 (1952) 525 1451 188 co54 c.f.cook and t.w. bonner,phys.rev. 94,651 (1954) 525 1451 189 co67 s.a. cox and f.r. pontet, j.nucl.energy 21,271 (1967) 525 1451 190 co69 j.a. cookson and j.g.locke,nucl.phys.a146,417(1970) 525 1451 191 co73 m.s. coates et al., priv. comm. to l.stewart (1973) 525 1451 192 da56 r.b.day,phys.rev.102,767 (1956) 525 1451 193 da60 r.b. day and m.walt,phys.rev.117,1330 (1960) 525 1451 194 da61 e.a. davis et al., nucl.phys.27,448 (1961) 525 1451 195 di67 k.m. diment, aere-r-5224 (1967) 525 1451 196 di88 j.k.dickens, proc.conf. on nuc.data for sci.& tech.,mito, 525 1451 197 japan (1988) p.213. 525 1451 198 fo61 d.m. fossan et al., phys.rev. 123,209 (1961) 525 1451 199 fr56 g.m. frye and j.h. gammel,phys.rev. 103,328 (1956) 525 1451 200 gl82 s.glendinning, nuc.sci.eng.80,256(1982) 525 1451 201 ha73 s.l.hausladen, thesis, ohio univ. coo-1717-5 (1973) 525 1451 202 he80 d.hetrick & c.y.fu, ornl/tm-7341 (1980). 525 1451 203 hy69 m.hyakutake, eandc(j)-13 (1969) p.29 525 1451 204 ho69 j.c. hopkins, priv. comm. lasl (1969) 525 1451 205 ir67 d.c.irving, ornl-tm-1872 (1967) 525 1451 206 ka87 R.Kavanagh & R.Marcley, Phys.Rev.C36, 1194 (1987). 525 1451 207 la71 r.o. lane et al., phys.rev.c4,380 (1971) 525 1451 208 ma68 r.l.macklin and j.h.gibbons,phys.rev.165,1147 (1968) 525 1451 209 mo66 f.p.mooring et al.,nucl.phys.82,16 (1966) 525 1451 210 ne54 n.g.nereson,la-1655 (1954) 525 1451 211 ne70 d.o.nellis et al., phys.rev. c1,847 (1970) 525 1451 212 po70 d.porter et al., awre o 45/70 (1970) 525 1451 213 qa85 S.Qaim et al., Santa Fe Conf. (1985)p.97. 525 1451 214 qa88 S.Qaim et al., Mito Conf. (1988) p.225. 525 1451 215 sa88 E.T. Sadowski, Ph.D thesis, Ohio U., (Nov.,1988). 525 1451 216 sc76 r.a. schrack et al., proc.icinn(erda-conf-760715-p2),1345 525 1451 217 (1976) 525 1451 218 se76 r.m. sealock and j.c. overley, phys.rev.c13,2149 (1976) 525 1451 219 si65 r.h.siemssen et al., nucl.phys.69,209 (1965) 525 1451 220 sp73 r.r. spencer et al., eandc(e)147,al (1973) 525 1451 221 te62 k.tesch, nucl.phys.37,412 (1962) 525 1451 222 th67 g.e. thomas et al., nucl.instr.meth.56,325 (1967) 525 1451 223 ts63 k.tsukada and o.tanaka,j.phys.soc.japan 18,610 (1963) 525 1451 224 va65 v.valkovic et al., phys.rev. 139,331 (1965) 525 1451 225 va70 b.vaucher et al.,helv.phys.acta 43,237 (1970) 525 1451 226 va72 l.van der zwan and k.w.geiger, nucl.phys. a180,615 (1972) 525 1451 227 wi55 h.b. willard et al., phys.rev. 98,669(1955) 525 1451 228 wy58 m.e. wyman et al., phys.rev.112,1264 (1958) 525 1451 229 **************************************************************** 525 1451 230 525 1451 231 525 1451 232 525 1451 233 525 1451 234 ***************** Program LINEAR (VERSION 2002-1) *************** 525 1451 235 For All Data Greater than 1.0000E-10 barns in Absolute Value 525 1451 236 Data Linearized to Within an Accuracy of .100000000 per-cent 525 1451 237 ***************** Program SIGMA1 (VERSION 2002-1) *************** 525 1451 238 Data Doppler Broadened to 300.000000 Kelvin 525 1451 239 for All Data Greater than 1.0000E-10 barns in Absolute Value 525 1451 240 Data Linearized to Within an Accuracy pf .100000000 per-cent 525 1451 241 ***************** Program FIXUP (Version 2002-1) **************** 525 1451 242 Corrected ZA/AWR in All Sections-----------------------------Yes 525 1451 243 Corrected Thresholds-----------------------------------------Yes 525 1451 244 Extended Cross Sections to 20 MeV----------------------------No 525 1451 245 Allow Cross Section Deletion---------------------------------No 525 1451 246 Allow Cross Section Reconstruction---------------------------No 525 1451 247 Make All Cross Sections Non-Negative-------------------------Yes 525 1451 248 Delete Energies Not in Ascending Order-----------------------Yes 525 1451 249 Deleted Duplicate Points-------------------------------------Yes 525 1451 250 Check for Ascending MAT/MF/MT Order--------------------------Yes 525 1451 251 Check for Legal MF/MT Numbers--------------------------------Yes 525 1451 252 Allow Creation of Missing Sections---------------------------No 525 1451 253 Allow Insertion of Energy Points-----------------------------No 525 1451 254 Create Uniform Energy Grid-----------------------------------No 525 1451 255 Delete Section if Cross Section =0 at All Energies-----------Yes 525 1451 256 **************************************************************** 525 1451 257 ***************** Program GROUPIE (VERSION 2002-1) ************** 525 1451 258 Unshielded Group Averages Using 640 Groups 525 1451 259 Weighting Spectrum: Flat (Constant) Spectrum 525 1451 260 1 451 266 2 525 1451 261 2 151 4 1 525 1451 262 3 107 217 1 525 1451 263 3 800 217 1 525 1451 264 3 801 217 1 525 1451 265 33 107 26 1 525 1451 266 525 1 099999 525 0 0 0 5.01000E+3 9.92692E+0 0 0 1 0 525 2151 1 5.010000+3 1.000000+0 0 0 1 0 525 2151 2 1.000000-5 1.000000+4 0 0 0 0 525 2151 3 3.000000+0 4.129038-1 0 0 0 0 525 2151 4 525 2 099999 525 0 0 0 5.01000E+3 9.92692E+0 0 0 0 0 525 3107 1 2.78950E+6 2.78950E+6 0 0 1 641 525 3107 2 641 1 525 3107 3 .000100000 60369.5672 .000105000 58961.8482 .000110000 57624.4985 525 3107 4 .000115000 56392.1633 .000120000 54951.5786 .000127500 53353.3087 525 3107 5 .000135000 51893.0043 .000142500 50551.0052 .000150000 49098.5863 525 3107 6 .000160000 47585.5417 .000170000 46208.8583 .000180000 44945.1556 525 3107 7 .000190000 43769.3848 .000200000 42695.9347 .000210000 41685.9908 525 3107 8 .000220000 40747.8940 .000230000 39876.3511 .000240000 38853.3697 525 3107 9 .000255000 37729.1547 .000270000 36861.6361 .000280000 35895.9179 525 3107 10 .000300000 34718.2100 .000320000 33651.8469 .000340000 32675.2286 525 3107 11 .000360000 31775.7003 .000380000 30954.5585 .000400000 30098.4665 525 3107 12 .000425000 29222.4439 .000450000 28422.1262 .000475000 27688.2832 525 3107 13 .000500000 26998.1136 .000525000 26367.3883 .000550000 25771.5792 525 3107 14 .000575000 25217.4640 .000600000 24649.5219 .000630000 24065.2488 525 3107 15 .000660000 23530.6294 .000690000 23017.7804 .000720000 22472.0914 525 3107 16 .000760000 21887.0079 .000800000 21343.7902 .000840000 20845.9067 525 3107 17 .000880000 20372.6916 .000920000 19938.9136 .000960000 19523.5195 525 3107 18 .001000000 19093.0153 .001050000 18643.5369 .001100000 18222.0987 525 3107 19 .001150000 17833.7845 .001200000 17375.7237 .001275000 16872.2935 525 3107 20 .001350000 16412.0235 .001425000 15982.4629 .001500000 15526.0689 525 3107 21 .001600000 15048.9035 .001700000 14613.2997 .001800000 14210.4042 525 3107 22 .001900000 13842.3306 .002000000 13501.1002 .002100000 13180.9304 525 3107 23 .002200000 12887.6835 .002300000 12607.2603 .002400000 12287.2985 525 3107 24 .002550000 11931.7648 .002700000 11653.6694 .002800000 11351.4747 525 3107 25 .003000000 10979.5222 .003200000 10641.0478 .003400000 10331.4803 525 3107 26 .003600000 10049.1544 .003800000 9788.49681 .004000000 9516.82460 525 3107 27 .004250000 9240.95277 .004500000 8989.16853 .004750000 8753.87133 525 3107 28 .005000000 8537.97874 .005250000 8337.88040 .005500000 8148.74821 525 3107 29 .005750000 7975.41761 .006000000 7793.39267 .006300000 7611.48177 525 3107 30 .006600000 7439.31120 .006900000 7279.58345 .007200000 7105.90399 525 3107 31 .007600000 6920.17898 .008000000 6750.36045 .008400000 6590.96048 525 3107 32 .008800000 6442.59109 .009200000 6304.75034 .009600000 6173.61303 525 3107 33 .010000000 6037.88045 .010500000 5894.40760 .011000000 5763.20486 525 3107 34 .011500000 5638.04017 .012000000 5494.58078 .012750000 5335.80072 525 3107 35 .013500000 5188.71615 .014250000 5053.72204 .015000000 4909.72079 525 3107 36 .016000001 4758.57103 .017000001 4619.99591 .017999999 4493.49187 525 3107 37 .018999999 4377.23066 .020000000 4268.17210 .021000000 4168.38441 525 3107 38 .022000000 4074.35803 .023000000 3986.73737 .024000000 3883.77635 525 3107 39 .025500000 3772.71905 .027000001 3684.72584 .028000001 3589.15197 525 3107 40 .029999999 3471.54644 .032000002 3364.41633 .034000002 3266.59939 525 3107 41 .035999998 3177.31827 .037999999 3094.94928 .039999999 3008.98714 525 3107 42 .042500000 2921.73869 .045000002 2841.98699 .047499999 2767.67893 525 3107 43 .050000001 2699.33694 .052499998 2636.10787 .055000000 2576.25246 525 3107 44 .057500001 2521.45255 .059999999 2463.84227 .063000001 2406.27149 525 3107 45 .066000000 2351.99591 .068999998 2301.39802 .071999997 2246.54223 525 3107 46 .075999998 2187.78934 .079999998 2134.10070 .083999999 2083.61392 525 3107 47 .088000000 2036.63468 .092000000 1993.11642 .096000001 1951.54781 525 3107 48 .100000001 1908.63090 .104999997 1863.20916 .109999999 1821.72938 525 3107 49 .115000002 1782.11180 .119999997 1736.73952 .127499998 1686.67539 525 3107 50 .135000005 1640.26788 .142499998 1597.62984 .150000006 1552.14713 525 3107 51 .159999996 1504.34851 .170000002 1460.63005 .180000007 1420.61485 525 3107 52 .189999998 1383.87368 .200000003 1349.37798 .209999993 1317.83677 525 3107 53 .219999999 1288.07328 .230000004 1260.35326 .239999995 1228.34986 525 3107 54 .254999995 1192.56069 .270000011 1165.17509 .280000001 1134.73256 525 3107 55 .300000012 1097.41711 .319999993 1063.54861 .340000004 1032.72552 525 3107 56 .360000014 1004.48688 .379999995 978.176957 .400000006 951.213540 525 3107 57 .425000012 923.728988 .449999988 898.318569 .474999994 874.931461 525 3107 58 .500000000 853.490362 .524999976 833.201485 .550000012 814.634634 525 3107 59 .574999988 796.994965 .600000024 778.993612 .629999995 760.736577 525 3107 60 .660000026 743.432752 .689999998 727.602537 .720000029 710.025968 525 3107 61 .759999990 691.623226 .800000012 674.579619 .839999974 658.588086 525 3107 62 .879999995 643.969952 .920000017 629.945238 .959999979 617.109609 525 3107 63 1.00000000 603.277354 1.04999995 589.134241 1.10000002 575.914847 525 3107 64 1.14999998 563.477910 1.20000005 549.191424 1.27499998 533.155511 525 3107 65 1.35000002 518.464556 1.42499995 505.068665 1.50000000 490.570975 525 3107 66 1.60000002 475.441752 1.70000005 461.668810 1.79999995 449.098119 525 3107 67 1.89999998 437.352915 2.00000000 426.522756 2.09999990 416.533362 525 3107 68 2.20000005 407.117410 2.29999995 398.487909 2.40000010 388.235804 525 3107 69 2.54999995 376.925420 2.70000005 368.281533 2.79999995 358.588483 525 3107 70 3.00000000 346.820994 3.20000005 336.155861 3.40000010 326.438320 525 3107 71 3.59999990 317.486077 3.79999995 309.372904 4.00000000 300.590352 525 3107 72 4.25000000 291.970516 4.50000000 284.063655 4.75000000 276.629867 525 3107 73 5.00000000 269.853908 5.25000000 263.336387 5.50000000 257.438448 525 3107 74 5.75000000 251.877635 6.00000000 246.184719 6.30000019 240.498371 525 3107 75 6.59999990 235.083594 6.90000010 229.877311 7.19999981 224.431765 525 3107 76 7.59999990 218.711104 8.00000000 213.172241 8.39999962 208.182068 525 3107 77 8.80000019 203.476760 9.19999981 199.106627 9.60000038 194.998681 525 3107 78 10.0000000 190.635152 10.5000000 186.176247 11.0000000 181.958221 525 3107 79 11.5000000 178.148941 12.0000000 173.502593 12.7500000 168.465281 525 3107 80 13.5000000 163.827889 14.2500000 159.535271 15.0000000 154.977384 525 3107 81 16.0000000 150.262102 17.0000000 145.825636 18.0000000 141.871954 525 3107 82 19.0000000 138.108916 20.0000000 134.675744 21.0000000 131.525133 525 3107 83 22.0000000 128.538764 23.0000000 125.790346 24.0000000 122.536877 525 3107 84 25.5000000 118.974574 27.0000000 116.301655 28.0000000 113.225799 525 3107 85 30.0000000 109.477352 32.0000000 106.090442 34.0000000 103.007267 525 3107 86 36.0000000 100.230060 38.0000000 97.5698278 40.0000000 94.8582851 525 3107 87 42.5000000 92.1159418 45.0000000 89.5814949 47.5000000 87.2222833 525 3107 88 50.0000000 85.0780678 52.5000000 83.0541719 55.0000000 81.1877474 525 3107 89 57.5000000 79.4434680 60.0000000 77.6493833 63.0000000 75.8552663 525 3107 90 66.0000000 74.1216311 69.0000000 72.5099356 72.0000000 70.8077030 525 3107 91 76.0000000 68.9334415 80.0000000 67.2309303 84.0000000 65.6683940 525 3107 92 88.0000000 64.1890143 92.0000000 62.7664324 96.0000000 61.4739442 525 3107 93 100.000000 60.1068063 105.000000 58.7169208 110.000000 57.3632577 525 3107 94 115.000000 56.1583737 120.000000 54.6872713 127.500000 53.0843480 525 3107 95 135.000000 51.6678067 142.500000 50.3016292 150.000000 48.8762241 525 3107 96 160.000000 47.3368294 170.000000 45.9692378 180.000000 44.6800160 525 3107 97 190.000000 43.5369068 200.000000 42.4336147 210.000000 41.4356567 525 3107 98 220.000000 40.4938801 230.000000 39.6289748 240.000000 38.6048846 525 3107 99 255.000000 37.4793437 270.000000 36.6321547 280.000000 35.6542795 525 3107 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4.365900-5 1.552481-5 1.322365-5 4.232958-5 2.020522-5 52533107 25 1.222438-4 52533107 26 52533 099999 525 0 0 0 0 0 0 0 9.01900E+3 1.88352E+1 0 0 34 10 925 1451 1 0.0 0.0 0 0 0 6 925 1451 2 1.00000E+0 3.00001E+7 0 0 10 2002 925 1451 3 3.00000E+2 0.0 1 0 164 4 925 1451 4 9-F - 19 FEI EVAL-OCT01 K.I.Zolotarev 925 1451 5 DIST-Feb2004 925 1451 6 ----IRDF-2002 MATERIAL 925 925 1451 7 -----INCIDENT NEUTRON DATA 925 1451 8 ------ENDF-6 FORMAT 925 1451 9 ***************************************************************** 925 1451 10 9-F -19 FEI EVAL-OCT01 K.I.Zolotarev 925 1451 11 DIST-JAN02 20020125 925 1451 12 ----BROND-2 MATERIAL 925 925 1451 13 -----INCIDENT NEUTRON DATA 925 1451 14 ------ENDF-6 FORMAT 925 1451 15 ------Russian Reactor Dosimetry File RRDF-2002 925 1451 16 ***************************************************************** 925 1451 17 Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko 925 1451 18 ***************************************************************** 925 1451 19 MF= 3 925 1451 20 MT= 16 -(n,2n) cross section data 925 1451 21 -------------------------------------- 925 1451 22 Excitation function for the F-19(n,2n)F-18 reaction in the 925 1451 23 energy region from threshold to 30 MeV was evaluated by means of 925 1451 24 statistical analysis of experimental cross section data [1-17] and 925 1451 25 data from GNASH [18,19] calculation. 925 1451 26 Experimental data [2-3],[5],[9-13],[15-17] were renormalized 925 1451 27 to the new standards for monitor reactions cross sections. Data 925 1451 28 of Mc Crary et al. [1], Rayburn et al.[2], Bormann et al.[4], and 925 1451 29 Shiokawa et al. [7] were renormalized to the results of precise 925 1451 30 measurements Vonach et al. [9] and Ikeda et al. [15] in the 14 MeV 925 1451 31 region. Data of Chatterjee et al. [8] were corrected to their pre- 925 1451 32 vious measurements at the 14.8 MeV point [6]. Results of measure- 925 1451 33 ments of Nagel [5] were taken only from SUBENT 005, EXFOR 20198. 925 1451 34 Experimental cross section data from ref. [20-31] were rejec- 925 1451 35 ted due to their big discrepancy with the main bulk of experimen- 925 1451 36 al data [1-17] and data from theoretical model calculation. 925 1451 37 The final procedure of evaluation F-19(n,2n)F-18 excitation 925 1451 38 function from threshold to 30 MeV has been carried out within the 925 1451 39 framework of generalized least squares method. Rational function 925 1451 40 was used as model function [32]. Calculations was performed by 925 1451 41 means of Pade-2 code [33]. 925 1451 42 U-235 thermal fission [34] and Cf-252 spontaneous fission 925 1451 43 neutron spectra [35] averaged cross-sections calculated from the 925 1451 44 evaluated F-19(n,2n)F-18 excitation function are the following: 925 1451 45 925 1451 46 ----------------------------------------------------------------- 925 1451 47 TYPE OF SPECTRUM I ,mb (calc.) I , mb (measured) 925 1451 48 ----------------------I-----------------I------------------------ 925 1451 49 U-235 neutron fission I 0.007299 I 0.007200+-0.001000 [36] 925 1451 50 I I 0.006509+-0.000300 [37] 925 1451 51 I I 0.008653+-0.000464 [38] 925 1451 52 ----------------------I-----------------I------------------------ 925 1451 53 CF-252 spont. fission I 0.01615 I 0.01628+-0.00054 [39] 925 1451 54 I I 0.01612+-0.00054 [40] 925 1451 55 ----------------------------------------------------------------- 925 1451 56 925 1451 57 MF=33 925 1451 58 MT= 16 -(n,2n) cross section cov. matrix 925 1451 59 --------------------------------------------- 925 1451 60 Uncertainties in the evaluated excitation function for the 925 1451 61 reaction F-19(n,2n)F-18 are given in the form of relative covari- 925 1451 62 ance matrix for the 26-neutron energy groups (LB=5). Covariance 925 1451 63 matrix of uncertainties was calculated simultaneously with 925 1451 64 recommended cross section data by means of PADE-2 code. 925 1451 65 Eigenvalues of the 6-th digits relative covariance matrix 925 1451 66 given in the 33-file are the following: 925 1451 67 925 1451 68 3.67622E-08 3.74071E-08 3.79051E-08 3.91111E-08 925 1451 69 4.11094E-08 4.32321E-08 4.68199E-08 4.98738E-08 925 1451 70 5.55426E-08 6.01322E-08 7.00100E-08 7.85485E-08 925 1451 71 9.57150E-08 1.06820E-07 1.28694E-07 1.47868E-07 925 1451 72 2.58858E-07 6.32301E-07 3.80384E-06 5.45023E-04 925 1451 73 1.36448E-03 3.28506E-03 8.26671E-03 1.39754E-02 925 1451 74 1.58981E-02 6.37025E-02 925 1451 75 925 1451 76 References : 925 1451 77 1. J.H.Mc Crary,I.L.Morgan Bull. American Phys. Soc., v.5, 925 1451 78 p.246, April 1960 ; 925 1451 79 J.H.Mc Crary,I.L.Morgan Report AFSWC-TR-60-30, 1960 925 1451 80 2. L.A.Rayburn Proc. of Conf. on Direct Interactions and Nuclear 925 1451 81 Reaction Mechanisms, Padua, 3-8 September 1962, Gordon and 925 1451 82 Breach, New York, 1963, p.322. ; 925 1451 83 L.A.Rayburn Bull. Am. Phys. Soc., v.7, p.335, April 1962 925 1451 84 3. M.Cevolani, S.Petralia Nuovo Cimento, v.26, p.1328, Dec.1962 925 1451 85 4. M.Bormann et al. Nucl. Phys., v.63, p.438, March 1965 925 1451 86 5. W.Nagel EXFOR 20198.005, 1966 925 1451 87 6. A.Chatterjee et al. Progress Report BARC-305, p.30, Nov. 1967 925 1451 88 7. T.Shiokawa et al. J. Inorg. Nucl. Chem., v.30, p.1, Jan. 1968 925 1451 89 8. A.Chatterjee e.a. Proc.of Symposium Nucl.Phys. and Solid State 925 1451 90 Phys., Roorke, India, 1969, v.2, p.117, December 1969 925 1451 91 9. H.K.Vonach et al. Proc. of 2-nd Conference on Nuclear Cross- 925 1451 92 Sections and Technology, Washington D.C., 4-7 March 1968, v.2, 925 1451 93 p.885 925 1451 94 10. R.C.Barrall et al. Report AFWL-TR-68-134, March 1969 925 1451 95 11. R.Mogharrab, H.Neuert Atomkernenergie, v.19, p.107, April 1972 925 1451 96 12. J.C.Robertson et al. J. Nucl. Energ., v.27, p.531, Aug. 1973 925 1451 97 13. R.A.Sigg Dissert. Abstr., sect. B, v.37, p.2237, Nov. 1976 925 1451 98 14. T.B.Ryves et al. J. of Physics, pt.G, v.4, n.11, p.1783, 1978 925 1451 99 15. Y.Ikeda et al. Report JAERI-1312, 1988 925 1451 100 16. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 925 1451 101 17. C.L.Hartmann, P.M.DeLuca Nucl. Sci. Eng., v.109, p.319, 1991 925 1451 102 18. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 925 1451 103 Model Code for Calculation of Cross Section and Emission 925 1451 104 Spectra. Report LA-6947, Los Alamos, 1977 925 1451 105 19. E.L.Trykov, G.Ya.Tertychnyi Private communication, IPPE, 925 1451 106 Obninsk, May 1999 925 1451 107 20. E.B.Paul, R.L.Clarke Canadian J. Phys., v.31, p.267, 1953 925 1451 108 21. V.J.Ashby e.a. Phys. Rev., v.111, p.616, 1958 925 1451 109 22. L.A.Rayburn Phys. Rev., v.122, p.168, 1961 925 1451 110 23. O.D.Brill et al. Doklady Akademii Nauk, v.136,n.1,p.55, 1961; 925 1451 111 O.D.Brill et al. Soviet Physics-Doklady, v.6, p.24, 1961 925 1451 112 24. J.Picard, C.F.Williamson Nucl. Phys., v.63, p.673, April 1965 925 1451 113 25. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 925 1451 114 26. J.Csikai Report EANDC-50 S, paper 102, July 1965 925 1451 115 27. A.Pasquarelli Nucl. Phys. A, v.93, p.218, March 1967 925 1451 116 28. M.Bormann, I.Riehle Zeitschrift f.Physik, v.207, p.64, 925 1451 117 October 1967 925 1451 118 29. H.O.Menlove et al. Phys. Rev., v.163, p.1308, 1967 925 1451 119 30. D.Crumpton J.Inorg. Nucl. Chem., v.31, p.3727, December 1969 925 1451 120 31. J.Araminowicz, J.Dresler Report INR-1464, p.14, May 1973 925 1451 121 32. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 925 1451 122 st's Meeting on Evaluation and Processing of Covariance Data, 925 1451 123 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 925 1451 124 33. S.A.Badikov et.al. Preprint FEI-1686, Obninsk, 1985 925 1451 125 34. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 925 1451 126 Library, MAT=9228, MF=5, MT=18, eval. April 1989 925 1451 127 35. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 925 1451 128 36. F.Nasyrov, B.D.Sciborskij Atomnaja Energija, v.25, no.5, 925 1451 129 p.437, November 1968 925 1451 130 37. M.Najzer, J.Rant Report IAEA-208, v.2, p.247, 1978 925 1451 131 38. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 925 1451 132 39. W.Mannhart Handbook on Nuclear Activation Data. IAEA Tech- 925 1451 133 nical Report series No.273, p.413, 1987 925 1451 134 40. W.Mannhart Validation of Differential Cross Sections with 925 1451 135 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 925 1451 136 September 2002 925 1451 137 ***************************************************************** 925 1451 138 File 2 added to the pointwise file containing only the effective 925 1451 139 scattering radius with no resonance parameters given. 925 1451 140 Taken from ENDF/B-VI 925 1451 141 925 1451 142 ***************************************************************** 925 1451 143 ***************** Program LINEAR (VERSION 2002-1) *************** 925 1451 144 For All Data Greater than 1.0000E-10 barns in Absolute Value 925 1451 145 Data Linearized to Within an Accuracy of .100000000 per-cent 925 1451 146 ***************** Program SIGMA1 (VERSION 2002-1) *************** 925 1451 147 Data Doppler Broadened to 300.000000 Kelvin 925 1451 148 for All Data Greater than 1.0000E-10 barns in Absolute Value 925 1451 149 Data Linearized to Within an Accuracy pf .100000000 per-cent 925 1451 150 ***************** Program FIXUP (Version 2002-1) **************** 925 1451 151 Corrected ZA/AWR in All Sections-----------------------------Yes 925 1451 152 Corrected Thresholds-----------------------------------------Yes 925 1451 153 Extended Cross Sections to 20 MeV----------------------------No 925 1451 154 Allow Cross Section Deletion---------------------------------No 925 1451 155 Allow Cross Section Reconstruction---------------------------No 925 1451 156 Make All Cross Sections Non-Negative-------------------------Yes 925 1451 157 Delete Energies Not in Ascending Order-----------------------Yes 925 1451 158 Deleted Duplicate Points-------------------------------------Yes 925 1451 159 Check for Ascending MAT/MF/MT Order--------------------------Yes 925 1451 160 Check for Legal MF/MT Numbers--------------------------------Yes 925 1451 161 Allow Creation of Missing Sections---------------------------No 925 1451 162 Allow Insertion of Energy Points-----------------------------No 925 1451 163 Create Uniform Energy Grid-----------------------------------No 925 1451 164 Delete Section if Cross Section =0 at All Energies-----------Yes 925 1451 165 ***************** Program GROUPIE (VERSION 2002-1) ************** 925 1451 166 Unshielded Group Averages Using 640 Groups 925 1451 167 Weighting Spectrum: Flat (Constant) Spectrum 925 1451 168 1 451 172 1 925 1451 169 2 151 4 1 925 1451 170 3 16 34 1 925 1451 171 33 16 71 1 925 1451 172 925 1 099999 925 0 0 0 9.01900E+3 1.88352E+1 0 0 1 0 925 2151 1 9.019000+3 1.000000+0 0 0 1 0 925 2151 2 1.000000-5 2.000000+7 0 0 0 0 925 2151 3 5.000000-1 5.360000-1 0 0 0 0 925 2151 4 925 2 099999 925 0 0 0 9.01900E+3 1.88352E+1 0 0 0 0 925 3 16 1 -1.04322E+7-1.04322E+7 0 0 1 92 925 3 16 2 92 1 925 3 16 3 10900000.0 2.02752E-6 11000000.0 5.63229E-5 11100000.0 .000137655 925 3 16 4 11200000.0 .000292275 11300000.0 .000548721 11400000.0 .000917543 925 3 16 5 11500000.0 .001402145 11600000.0 .002002095 11700000.0 .002714490 925 3 16 6 11800000.0 .003534705 11900000.0 .004456885 12000000.0 .005474310 925 3 16 7 12100000.0 .006579685 12200000.0 .007765400 12300000.0 .009023740 925 3 16 8 12400000.0 .010347050 12500000.0 .011727850 12600000.0 .013159000 925 3 16 9 12700000.0 .014633700 12800000.0 .016152650 12900000.0 .017695750 925 3 16 10 13000000.0 .019265950 13100000.0 .020863250 13200000.0 .022460550 925 3 16 11 13300000.0 .024071763 13400000.0 .025696888 13500000.0 .027322013 925 3 16 12 13600000.0 .028947138 13700000.0 .030572263 13800000.0 .032197388 925 3 16 13 13900000.0 .033822513 14000000.0 .035447638 14100000.0 .037049625 925 3 16 14 14200000.0 .038628475 14300000.0 .040207325 14400000.0 .041786175 925 3 16 15 14500000.0 .043331888 14600000.0 .044844463 14700000.0 .046357038 925 3 16 16 14800000.0 .047869613 14900000.0 .049339675 15000000.0 .050767225 925 3 16 17 15100000.0 .052194775 15200000.0 .053622325 15300000.0 .054999450 925 3 16 18 15400000.0 .056326150 15500000.0 .057652850 15600000.0 .058979550 925 3 16 19 15700000.0 .060248850 15800000.0 .061460750 15900000.0 .062672650 925 3 16 20 16000000.0 .063884550 16100000.0 .065033000 16200000.0 .066118000 925 3 16 21 16300000.0 .067203000 16400000.0 .068288000 16500000.0 .069304550 925 3 16 22 16600000.0 .070252650 16700000.0 .071200750 16800000.0 .072148850 925 3 16 23 16900000.0 .073024838 17000000.0 .073828713 17100000.0 .074632588 925 3 16 24 17200000.0 .075436463 17300000.0 .076166188 17400000.0 .076821763 925 3 16 25 17500000.0 .077477338 17600000.0 .078132913 17700000.0 .078714000 925 3 16 26 17800000.0 .079220600 17900000.0 .079727200 18000000.0 .080233800 925 3 16 27 18100000.0 .080667413 18200000.0 .081028038 18300000.0 .081388663 925 3 16 28 18400000.0 .081749288 18500000.0 .082040025 18600000.0 .082260875 925 3 16 29 18700000.0 .082481725 18800000.0 .082702575 18900000.0 .082858125 925 3 16 30 19000000.0 .082948375 19100000.0 .083038625 19200000.0 .083128875 925 3 16 31 19300000.0 .083180900 19400000.0 .083152770 19500000.0 .083082710 925 3 16 32 19600000.0 .083012650 19700000.0 .082942590 19800000.0 .082872530 925 3 16 33 19900000.0 .082740370 20000000.0 0.0 925 3 16 34 925 3 099999 925 0 0 0 9.01900E+3 1.88352E+1 0 0 0 1 92533 16 1 0.000000+0 0.000000+0 0 16 0 1 92533 16 2 0.000000+0 0.000000+0 1 5 406 28 92533 16 3 1.000000-5 1.090000+7 1.200000+7 1.250000+7 1.300000+7 1.350000+7 92533 16 4 1.400000+7 1.450000+7 1.500000+7 1.550000+7 1.600000+7 1.650000+7 92533 16 5 1.700000+7 1.750000+7 1.800000+7 1.850000+7 1.900000+7 1.950000+7 92533 16 6 2.000000+7 2.100000+7 2.200000+7 2.300000+7 2.400000+7 2.500000+7 92533 16 7 2.700000+7 2.800000+7 2.900000+7 3.000000+7 0.000000+0 0.000000+0 92533 16 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 92533 16 9 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 92533 16 10 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 92533 16 11 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 92533 16 12 0.000000+0 7.844560-3 4.898350-3 3.027060-3 1.929170-3 1.321910-3 92533 16 13 9.926111-4 8.097050-4 6.990610-4 6.218640-4 5.596050-4 5.048120-4 92533 16 14 4.555230-4 4.121780-4 3.758470-4 3.473540-4 3.269110-4 3.140740-4 92533 16 15 3.073570-4 3.120960-4 3.231580-4 3.310920-4 3.298620-4 3.041000-4 92533 16 16 2.525070-4 2.039030-4 1.464150-4 3.648070-3 2.360540-3 1.449110-3 92533 16 17 9.104930-4 6.319380-4 5.059420-4 4.569550-4 4.388680-4 4.266530-4 92533 16 18 4.088340-4 3.819840-4 3.470900-4 3.071470-4 2.656870-4 2.259100-4 92533 16 19 1.902720-4 1.486420-4 1.141580-4 1.016320-4 1.039070-4 1.134800-4 92533 16 20 1.281520-4 1.350840-4 1.314670-4 1.212070-4 1.826270-3 1.354210-3 92533 16 21 9.782671-4 7.059900-4 5.267220-4 4.207220-4 3.665780-4 3.449020-4 92533 16 22 3.398810-4 3.397380-4 3.365240-4 3.256910-4 3.054490-4 2.760780-4 92533 16 23 2.392420-4 1.754770-4 8.900070-5 1.611170-5-3.205140-5-5.099350-5 92533 16 24 -2.365820-5 5.330250-5 1.291950-4 2.188860-4 1.233180-3 1.050650-3 92533 16 25 8.482810-4 6.625510-4 5.129570-4 4.047920-4 3.347390-4 2.952800-4 92533 16 26 2.776250-4 2.733860-4 2.755370-4 2.787900-4 2.796490-4 2.762050-4 92533 16 27 2.614270-4 2.288330-4 1.889520-4 1.513570-4 1.231830-4 1.085890-4 92533 16 28 1.236560-4 1.520640-4 1.922260-4 1.030240-3 9.309550-4 7.910340-4 92533 16 29 6.422140-4 5.056870-4 3.931520-4 3.093600-4 2.543630-4 2.252570-4 92533 16 30 2.174790-4 2.257800-4 2.449250-4 2.701550-4 3.105180-4 3.558030-4 92533 16 31 3.811660-4 3.836510-4 3.655150-4 3.091160-4 2.316670-4 1.739440-4 92533 16 32 1.144030-4 9.301140-4 8.611210-4 7.506000-4 6.229440-4 4.974770-4 92533 16 33 3.876710-4 3.015590-4 2.426370-4 2.107830-4 2.032840-4 2.157690-4 92533 16 34 2.430530-4 3.003160-4 3.817500-4 4.465680-4 4.803090-4 4.784630-4 92533 16 35 4.115300-4 2.889300-4 1.821310-4 6.207860-5 8.648120-4 8.135850-4 92533 16 36 7.261930-4 6.210140-4 5.137590-4 4.162000-4 3.360620-4 2.773020-4 92533 16 37 2.407400-4 2.248090-4 2.263440-4 2.532360-4 3.094430-4 3.658040-4 92533 16 38 4.035410-4 4.133410-4 3.692840-4 2.697090-4 1.753330-4 6.488820-5 92533 16 39 8.276580-4 7.994630-4 7.403910-4 6.622230-4 5.756870-4 4.895590-4 92533 16 40 4.102830-4 3.420170-4 2.868660-4 2.452630-4 2.075860-4 1.890180-4 92533 16 41 1.937990-4 2.070150-4 2.180840-4 2.167570-4 1.920930-4 1.606910-4 92533 16 42 1.194350-4 8.367870-4 8.381600-4 8.077220-4 7.513760-4 6.760250-4 92533 16 43 5.886680-4 4.958200-4 4.031020-4 3.150170-4 2.001340-4 8.278800-5 92533 16 44 1.074360-5-2.078540-5-2.009100-5 2.343720-5 9.478700-5 1.508320-4 92533 16 45 2.094530-4 9.030970-4 9.293760-4 9.163470-4 8.669560-4 7.869960-4 92533 16 46 6.839560-4 5.660230-4 4.412210-4 2.582200-4 4.586960-5-1.068070-4 92533 16 47 -1.912730-4-2.107930-4-1.358110-4 2.138250-5 1.612100-4 3.183170-4 92533 16 48 1.012600-3 1.049310-3 1.037800-3 9.812610-4 8.866920-4 7.633760-4 92533 16 49 6.215440-4 3.970180-4 1.170730-4-1.006010-4-2.340860-4-2.800390-4 92533 16 50 -1.974460-4 1.093280-5 2.077500-4 4.350440-4 1.136340-3 1.170610-3 92533 16 51 1.152160-3 1.086070-3 9.810220-4 8.476920-4 6.195350-4 3.158060-4 92533 16 52 6.329550-5-1.052330-4-1.794050-4-1.175480-4 9.273780-5 3.048660-4 92533 16 53 5.562250-4 1.254720-3 1.286290-3 1.267530-3 1.204830-3 1.107390-3 92533 16 54 9.184080-4 6.431870-4 3.946850-4 2.121830-4 1.124980-4 1.261560-4 92533 16 55 2.843660-4 4.638940-4 6.851270-4 1.376450-3 1.420770-3 1.422120-3 92533 16 56 1.386320-3 1.278440-3 1.085040-3 8.822720-4 7.106800-4 5.921990-4 92533 16 57 5.353610-4 5.914810-4 6.921020-4 8.298950-4 1.539830-3 1.622460-3 92533 16 58 1.669850-3 1.679580-3 1.616300-3 1.498950-3 1.363950-3 1.236560-3 92533 16 59 1.094900-3 1.009270-3 9.922760-4 1.001160-3 1.798240-3 1.944450-3 92533 16 60 2.099870-3 2.205700-3 2.207720-3 2.133310-3 2.009010-3 1.777510-3 92533 16 61 1.524740-3 1.362900-3 1.209860-3 2.198710-3 2.517890-3 2.820120-3 92533 16 62 2.967440-3 2.974120-3 2.866070-3 2.549010-3 2.119470-3 1.798830-3 92533 16 63 1.465550-3 3.093180-3 3.712880-3 4.106620-3 4.263970-3 4.206430-3 92533 16 64 3.790030-3 3.114020-3 2.560510-3 1.957740-3 4.741660-3 5.470710-3 92533 16 65 5.854090-3 5.903210-3 5.430360-3 4.514070-3 3.710880-3 2.809610-3 92533 16 66 6.497010-3 7.107000-3 7.299680-3 6.881850-3 5.882950-3 4.952680-3 92533 16 67 3.884000-3 7.920590-3 8.280630-3 8.031330-3 7.128890-3 6.218540-3 92533 16 68 5.143690-3 8.820440-3 8.840330-3 8.205440-3 7.459920-3 6.540140-3 92533 16 69 9.408330-3 9.443150-3 9.180490-3 8.759390-3 1.037760-2 1.080020-2 92533 16 70 1.110710-2 1.176100-2 1.264710-2 1.416170-2 92533 16 71 92533 099999 925 0 0 0 0 0 0 0 1.10230E+4 2.27923E+1 0 0 34 101125 1451 1 0.0 0.0 0 0 0 61125 1451 2 1.00000E+0 2.00000E+7 0 0 10 20021125 1451 3 3.00000E+2 0.0 1 0 214 61125 1451 4 11-Na- 23 ORNL,TOH EVAL-DEC77 D. C. LARSON,S.IWASAKI 1125 1451 5 DIST-Feb2004 1125 1451 6 ----IRDF-2002 MATERIAL 1125 1125 1451 7 -----INCIDENT NEUTRON DATA 1125 1451 8 ------ENDF-6 FORMAT 1125 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************1125 1451 10 DOSIMETRY 1125 1451 11 ******************************************************************1125 1451 12 11-NA- 23 ORNL EVAL-DEC77 D. C. LARSON 1125 1451 13 DIST-SEP 1 REV1-JUL91 20010926 1125 1451 14 ----ENDF/B-VI MATERIAL 1125 REVISION 1 1125 1451 15 1125 1451 16 **************************************************************** 1125 1451 17 11-NA- 23 TOH EVAL-AUG96 S.IWASAKI 1125 1451 18 DIST-JUL98 1125 1451 19 ----JENDL/D-99 MATERIAL 1125 1125 1451 20 -----INCIDENT NEUTRON DATA 1125 1451 21 ------ENDF-6 FORMAT 1125 1451 22 1125 1451 23 NA-23 (N,2N) NA-22 (HALF-LIFE = 2.6019Y) 1125 1451 24 **************************************************************** 1125 1451 25 ================================================================ 1125 1451 26 ************** Start of ENDF/B-VI Bibliography ************** 1125 1451 27 ================================================================ 1125 1451 28 ENDF/B-VI MOD 2 Revision, September 2000, S.C.Frankle, R.C.Reedy,1125 1451 29 P.G.Young (LANL) 1125 1451 30 1125 1451 31 The secondary gamma-ray spectrum for radiative capture (MF 12, 1125 1451 32 MT 102) has been updated using new experimental data. 1125 1451 33 The previous evaluation (MOD 1: 61 gamma rays) is replaced with 1125 1451 34 292 discrete gamma rays based on thermal neutron experimental 1125 1451 35 data. Thermal spectrum assumed to 20 MeV. 1125 1451 36 The Q-value for radiative capture was also updated in File 3. 1125 1451 37 1125 1451 38 Gas production added as MF=3, MT=203,207 1125 1451 39 1125 1451 40 Details of these changes are described in Frankle et al. [Fr01]. 1125 1451 41 1125 1451 42 **************************************************************** 1125 1451 43 1125 1451 44 ENDF/B-VI MOD 1 Revision, July 1991, V. McLane (NNDC) 1125 1451 45 1125 1451 46 REVISION 1 CORRECTION 1125 1451 47 1125 1451 48 Corrected number of energy ranges in File 32, MT = 151 to be 1. 1125 1451 49 1125 1451 50 **************************************************************** 1125 1451 51 1125 1451 52 ENDF/B-VI MOD 0, January 1990 (NNDC) 1125 1451 53 1125 1451 54 ENDF/B-V Material 1311 converted to ENDF-6 format by NNDC. 1125 1451 55 1125 1451 56 **************************************************************** 1125 1451 57 1125 1451 58 ENDF/B-V Evaluation, December 1977, D. C. Larson (ORNL) 1125 1451 59 1125 1451 60 MF=2 1125 1451 61 Resonance parameters are used from 600 eV to 500 keV. The 1125 1451 62 thermal capture cross section is given as 528 mb, in agreement 1125 1451 63 with the experimental results of 526.9+-4.5 mb [RY71], as well 1125 1451 64 as earlier results summarized in [RY71]. Using resonance 1125 1451 65 parameters E=2.81 keV, GN=376 eV, and GG=0.353 eV for the large 1125 1451 66 2.81 keV resonance gives the correct thermal capture value using 1125 1451 67 a Breit-Wigner shape, and this form is used to calculate the 1125 1451 68 capture crposs section from 1.0E-5 to 600 eV. The capture 1125 1451 69 width GG=0.353 eV is consistent with a recent result by Wilson 1125 1451 70 [WI77] where they find 0.24 1125 1451 154 NA-23 (N,2N) NA-22 (HALF-LIFE = 2.6019Y) 1125 1451 155 1125 1451 156 MF=1 GENERAL INFORMATION 1125 1451 157 MT=451 DESCRIPTIVE DATA AND DICTIONARY 1125 1451 158 1125 1451 159 MF=3 NEUTRON CROSS SECTIONS 1125 1451 160 MT=16 (N,2N) CROSS SECTION 1125 1451 161 EVALUATED USING SPLINE FITTING METHOD /6/TO A SET OF 1125 1451 162 SELECTED EXPERIMENTAL DATA /9-14/ TAKEN FROM NESTOR-2/8/. 1125 1451 163 MAINLY BASED ON THE EXPERIMENTAL DATA OF ADAMSKI/6/. 1125 1451 164 1125 1451 165 MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS 1125 1451 166 MT=16 1125 1451 167 GENERATED USING SPLINE FITTING METHOD /6/ 1125 1451 168 1125 1451 169 REFERENCES 1125 1451 170 6) S. IWASAKI, NUCLEAR DATA FOR SCINECE AND TECHNOLOGY, 1125 1451 171 GATLINBURG,TENNESSEE, MAY 9 13, 1994. PROCEEDINGS, VOL 2, 614,1125 1451 172 ANS, INC., (1994). 1125 1451 173 8) T. NAKAGAWA : THE JAERI NUCLEAR DATA CENTER, UNPUBLISHED. 1125 1451 174 9) H.O. MENLOVE, ET AL. PHYS. REV. 163,1308 (1967). 1125 1451 175 10) L. ADAMSKI, ET AL. ANN. NUCL. ENERGY, 7, (7), 397, (1980). 1125 1451 176 11) Y. IKEDA, ET AL., JAERI-1312 (1988). 1125 1451 177 12) LU HAN-LIN , ET AL. INDC(CRP)-16 (1989). 1125 1451 178 13) B.STROHMAIER, ET AL. ANN. NUCL. ENERG., 20,8 , 533-545 (1991).1125 1451 179 14) M. SAKUMA, S. IWASAKI, H. SHIMADA, N. ODANO, K. SUDA, 1125 1451 180 J.R. DUMAIS,AND K. SUGIYAMA, JAERI-M 92-027, P. 278 (1992). 1125 1451 181 ================================================================ 1125 1451 182 ******* Processing history of JENDL/D99 (n,2n) component ******* 1125 1451 183 ================================================================ 1125 1451 184 ***************** Program LINEAR (VERSION 2002-1) ***************1125 1451 185 For All Data Greater than 1.0000E-10 barns in Absolute Value 1125 1451 186 Data Linearized to Within an Accuracy of .100000000 per-cent 1125 1451 187 ***************** Program SIGMA1 (VERSION 2002-1) ***************1125 1451 188 Data Doppler Broadened to 300.000000 Kelvin 1125 1451 189 for All Data Greater than 1.0000E-10 barns in Absolute Value 1125 1451 190 Data Linearized to Within an Accuracy pf .100000000 per-cent 1125 1451 191 ***************** Program FIXUP (Version 2002-1) ****************1125 1451 192 Corrected ZA/AWR in All Sections-----------------------------Yes 1125 1451 193 Corrected Thresholds-----------------------------------------Yes 1125 1451 194 Extended Cross Sections to 20 MeV----------------------------No 1125 1451 195 Allow Cross Section Deletion---------------------------------No 1125 1451 196 Allow Cross Section Reconstruction---------------------------No 1125 1451 197 Make All Cross Sections Non-Negative-------------------------Yes 1125 1451 198 Delete Energies Not in Ascending Order-----------------------Yes 1125 1451 199 Deleted Duplicate Points-------------------------------------Yes 1125 1451 200 Check for Ascending MAT/MF/MT Order--------------------------Yes 1125 1451 201 Check for Legal MF/MT Numbers--------------------------------Yes 1125 1451 202 Allow Creation of Missing Sections---------------------------No 1125 1451 203 Allow Insertion of Energy Points-----------------------------No 1125 1451 204 Create Uniform Energy Grid-----------------------------------No 1125 1451 205 Delete Section if Cross Section =0 at All Energies-----------Yes 1125 1451 206 ================================================================ 1125 1451 207 *********** End of JENDL/D-99 (n,2n) Bibliography *********** 1125 1451 208 ================================================================ 1125 1451 209 1125 1451 210 1125 1451 211 1125 1451 212 1125 1451 213 ***************************************************************** 1125 1451 214 ************************ C O N T E N T S *********************** 1125 1451 215 ***************** Program GROUPIE (VERSION 2002-1) **************1125 1451 216 Unshielded Group Averages Using 640 Groups 1125 1451 217 Weighting Spectrum: Flat (Constant) Spectrum 1125 1451 218 1 451 224 21125 1451 219 2 151 4 01125 1451 220 3 16 27 11125 1451 221 3 102 217 21125 1451 222 33 16 32 11125 1451 223 33 102 5 01125 1451 224 1125 1 099999 1125 0 0 0 1.10230E+4 2.27922E+1 0 0 1 01125 2151 1 1.10230E+4 1.00000E+0 0 0 1 01125 2151 2 6.00000E+2 5.00000E+5 0 0 0 01125 2151 3 1.50000E+0 5.41000E-1 0 0 0 01125 2151 4 1125 2 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2.29330- 4 1.79550- 4 1.02600- 4 1.02600- 4 6.84000- 5112533 16 28 2.89000- 4 2.83220- 4 2.64180- 4 1.65240- 4 1.37700- 4 1.22400- 4112533 16 29 2.89000- 4 3.53430- 4 2.23380- 4 2.57040- 4 1.83600- 4 4.41000- 4112533 16 30 3.70440- 4 4.87620- 4 4.00680- 4 3.24000- 4 4.81140- 4 4.66560- 4112533 16 31 7.29000- 4 9.42840- 4 1.29600- 3 112533 16 32 112533 099999 1.10230E+4 2.27922E+1 0 0 0 1112533102 1 0.000000+0 0.000000+0 0 102 0 1112533102 2 0.000000+0 0.000000+0 0 1 12 6112533102 3 1.000000-5 4.000000-4 5.000000+1 2.500000-3 6.000000+2 1.000000-2112533102 4 5.000000+5 4.000000-2 5.000000+6 6.250000-2 2.000000+7 0.000000+0112533102 5 112533 099999 1125 0 0 0 0 0 0 0 1.20240E+4 2.37790E+1 0 0 34 101225 1451 1 0.0 0.0 0 0 0 61225 1451 2 1.00000E+0 2.10000E+7 0 0 10 20021225 1451 3 3.00000E+2 0.0 1 0 40 41225 1451 4 12-Mg- 24 IRK-VIENNA EVAL-APR90 1225 1451 5 DIST-Feb2004 1225 1451 6 ----IRDF-2002 MATERIAL 1225 1225 1451 7 -----INCIDENT NEUTRON DATA 1225 1451 8 ------ENDF-6 FORMAT 1225 1451 9 *****************************************************************1225 1451 10 12-MG- 24 IRK-VIENNA EVAL-APR90 1225 1451 11 DIST-JUN90 1225 1451 12 IRK-EVAL.NLIB 25 1225 1225 1451 13 *****************************************************************1225 1451 14 The Q values and threshold energies were updated prior to pro- 1225 1451 15 cessing through the codes to comply with the values obtained 1225 1451 16 using the NNDC calculation program which is based on the 1995 1225 1451 17 Update to the Atomic mass Evaluation. 1225 1451 18 *****************************************************************11225 1451 19 ***************** Program LINEAR (VERSION 2002-1) ***************1225 1451 20 For All Data Greater than 1.0000E-10 barns in Absolute Value 1225 1451 21 Data Linearized to Within an Accuracy of .100000000 per-cent 1225 1451 22 ***************** Program SIGMA1 (VERSION 2002-1) ***************1225 1451 23 Data Doppler Broadened to 300.000000 Kelvin 1225 1451 24 for All Data Greater than 1.0000E-10 barns in Absolute Value 1225 1451 25 Data Linearized to Within an Accuracy pf .100000000 per-cent 1225 1451 26 ***************** Program FIXUP (Version 2002-1) ****************1225 1451 27 Corrected ZA/AWR in All Sections-----------------------------Yes 1225 1451 28 Corrected Thresholds-----------------------------------------Yes 1225 1451 29 Extended Cross Sections to 20 MeV----------------------------No 1225 1451 30 Allow Cross Section Deletion---------------------------------No 1225 1451 31 Allow Cross Section Reconstruction---------------------------No 1225 1451 32 Make All Cross Sections Non-Negative-------------------------Yes 1225 1451 33 Delete Energies Not in Ascending Order-----------------------Yes 1225 1451 34 Deleted Duplicate Points-------------------------------------Yes 1225 1451 35 Check for Ascending MAT/MF/MT Order--------------------------Yes 1225 1451 36 Check for Legal MF/MT Numbers--------------------------------Yes 1225 1451 37 Allow Creation of Missing Sections---------------------------No 1225 1451 38 Allow Insertion of Energy Points-----------------------------No 1225 1451 39 Create Uniform Energy Grid-----------------------------------No 1225 1451 40 Delete Section if Cross Section =0 at All Energies-----------Yes 1225 1451 41 ***************** Program GROUPIE (VERSION 2002-1) **************1225 1451 42 Unshielded Group Averages Using 640 Groups 1225 1451 43 Weighting Spectrum: Flat (Constant) Spectrum 1225 1451 44 1 451 48 11225 1451 45 2 151 4 11225 1451 46 3 103 54 11225 1451 47 33 103 191 11225 1451 48 1225 1 099999 1225 0 0 0 1.20240E+4 2.37790E+1 0 0 1 01225 2151 1 1.20240E+4 1.00000E+0 0 0 1 01225 2151 2 1.00000E+0 2.00000E+7 0 0 0 01225 2151 3 2.50000E+0 3.15000E-1 0 0 0 01225 2151 4 1225 2 099999 1225 0 0 0 1.20240E+4 2.37790E+1 0 0 0 01225 3103 1 -4.73343E+6-4.73343E+6 0 0 1 1521225 3103 2 152 1 1225 3103 3 4900000.00 1.93283E-5 5000000.00 3.23719E-5 5100000.00 3.45514E-51225 3103 4 5200000.00 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4.102000-5 3.878000-5 1.003000-4 1.366000-4 7.423000-4122533103 156 3.689000-4 1.499000-4 1.471000-4 1.234000-4 6.027000-5 7.013000-5122533103 157 7.715000-6 7.101000-6 6.377000-7 6.900000-6 8.037000-6 9.057000-6122533103 158 7.065000-5 9.992000-5 7.427000-5 6.883000-5 5.504000-5 1.462000-4122533103 159 2.198000-4 0.000000+0 3.917000-4 1.415000-4 1.163000-4 6.212000-5122533103 160 6.163000-5 9.348000-6 5.029000-6 1.392000-6 4.268000-6 6.262000-6122533103 161 6.583000-6 4.990000-5 6.943000-5 5.552000-5 4.795000-5 3.995000-5122533103 162 1.171000-4 1.481000-4 0.000000+0 6.314000-4 7.655000-5 6.174000-5122533103 163 5.043000-5 1.095000-5 6.171000-6 2.942000-6 8.958000-7 7.855000-6122533103 164 2.145000-5 4.271000-5 5.936000-5 4.065000-5 4.325000-5 3.943000-5122533103 165 5.687000-5 0.000000+0 0.000000+0 4.217000-4 5.631000-5 6.805000-5122533103 166 9.824000-6 6.191000-6 4.993000-6 1.063000-5 1.087000-5 2.519000-5122533103 167 8.153000-5 1.136000-4 9.074000-5 7.833000-5 6.488000-5 1.916000-4122533103 168 2.484000-4 0.000000+0 1.364000-4 2.742000-5 5.675000-6 4.078000-6122533103 169 2.047000-6 3.265000-6 6.286000-6 7.306000-6 3.756000-5 3.566000-5122533103 170 3.666000-5 3.988000-5 3.708000-5 6.163000-5 8.552000-5 1.954000-4122533103 171 2.193000-4 4.042000-5 3.101000-5 3.731000-5 4.210000-5 2.190000-5122533103 172 6.633000-6 2.502000-5 3.546000-5 2.793000-5 2.402000-5 1.868000-5122533103 173 5.986000-5 9.835000-5 0.000000+0 2.990000-5 1.234000-5 1.606000-5122533103 174 1.677000-5 8.438000-6 2.638000-6 2.727000-6 4.223000-6 3.176000-6122533103 175 5.216000-6 5.098000-6 1.078000-5 1.432000-5 7.229000-5 2.146000-5122533103 176 1.217000-5 1.304000-5 7.310000-6 1.286000-6 1.058000-5 1.034000-5122533103 177 1.053000-5 9.525000-6 8.492000-6 1.116000-5 1.378000-5 0.000000+0122533103 178 2.728000-5 1.661000-5 8.260000-6 1.834000-6 1.729000-6 2.294000-6122533103 179 1.904000-6 3.962000-6 4.040000-6 9.790000-6 1.247000-5 6.928000-5122533103 180 3.051000-5 8.964000-6 1.622000-6 5.904000-6 8.294000-6 5.994000-6122533103 181 8.828000-6 7.746000-6 2.092000-5 3.809000-5 9.510000-5 3.034000-5122533103 182 1.586000-5 1.302000-5 1.255000-5 1.343000-5 1.401000-5 1.257000-5122533103 183 2.020000-5 2.765000-5 6.808000-5 1.409000-4 1.474000-5 2.281000-5122533103 184 1.493000-5 1.676000-5 1.412000-5 2.123000-5 1.232000-5 2.577000-5122533103 185 4.629000-4 1.240000-4 1.180000-4 1.074000-4 9.052000-5 1.470000-4122533103 186 1.828000-4 0.000000+0 4.242000-4 1.303000-4 1.194000-4 9.884000-5122533103 187 2.045000-4 2.627000-4 0.000000+0 4.462000-4 1.101000-4 9.018000-5122533103 188 1.636000-4 2.041000-4 0.000000+0 3.819000-4 1.030000-4 1.966000-4122533103 189 2.921000-4 6.338000-4 3.281000-4 1.682000-4 2.465000-4 5.976000-4122533103 190 1.025000-3 7.385000-4 1.611000-3 2.111000-3 3.391000-3 2.720000-2122533103 191 122533 099999 1225 0 0 0 0 0 0 0 1.30270E+4 2.67497E+1 0 0 34 101325 1451 1 0.0 0.0 0 0 0 61325 1451 2 1.00000E+0 2.00000E+7 0 0 10 20021325 1451 3 3.00000E+2 0.0 1 0 325 61325 1451 4 13-Al- 27 FEI/IRK EVAL-APR90 K.I.Zolotarev, M.Wagner et al. 1325 1451 5 DIST-Feb2004 1325 1451 6 ----IRDF-2002 MATERIAL 1325 1325 1451 7 -----INCIDENT NEUTRON DATA 1325 1451 8 ------ENDF-6 FORMAT 1325 1451 9 *****************************************************************1325 1451 10 13-AL- 27 FEI EVAL-May03 K.I.Zolotarev 1325 1451 11 DIST-Sep03 1325 1451 12 ----BROND-3 MATERIAL 1325 1325 1451 13 -----INCIDENT NEUTRON DATA 1325 1451 14 ------ENDF-6 FORMAT 1325 1451 15 ***************************************************************** 1325 1451 16 ------Russian Reactor Dosimetry File RRDF-2002 1325 1451 17 ***************************************************************** 1325 1451 18 Authors of evaluation: K.Zolotarev, A.Pashchenko, J.Csikai 1325 1451 19 ***************************************************************** 1325 1451 20 MF=3 1325 1451 21 MT=103 - (n,p) cross section 1325 1451 22 *****************************************************************1325 1451 23 13-AL- 27 IRK-VIENNA EVAL-APR90 1325 1451 24 DIST-JUN90 1325 1451 25 IRK-EVAL.NLIB 25 1325 1325 1451 26 MF=3 1325 1451 27 MT=107 - (n,a) cross section 1325 1451 28 ******************************************************************1325 1451 29 ***************************************************************** 1325 1451 30 ********** Start of (N,P) bibliographical component ********** 1325 1451 31 ***************************************************************** 1325 1451 32 ------Russian Reactor Dosimetry File RRDF-2002 1325 1451 33 ***************************************************************** 1325 1451 34 Authors of evaluation: K.Zolotarev, A.Pashchenko, J.Csikai 1325 1451 35 ***************************************************************** 1325 1451 36 MF=3 1325 1451 37 MT=103 - (n,p) cross section 1325 1451 38 ------------------------------------- 1325 1451 39 Microscopic experimental data [1-76] were analyzed in the 1325 1451 40 process of preparation of input data base for the evaluation of 1325 1451 41 cross sections and their uncertainty for the Al-27(n,p)Mg-27 1325 1451 42 reaction. During this procedure all experimental data if it was 1325 1451 43 possible were corrected to the new recommended cross section data 1325 1451 44 for monitor reactions used in the measurements and to the new re- 1325 1451 45 commended decay data from ref. [77]. 1325 1451 46 Excitation function for the Al-27(n,p)Mg-27 reaction in the 1325 1451 47 energy region from threshold to 23.0 MeV was evaluated by means 1325 1451 48 of statistical analysis of experimental cross section data [1-58].1325 1451 49 Special correction was done with experimental data [3], [8], 1325 1451 50 [12], [13], [15], [17], [27], [28], [47] and [58]. 1325 1451 51 Experimental data of Hudson and Morgan [3], Gabbard et al.[8],1325 1451 52 Ferguson and Albergotti [15] , Cuzzocrea et al. [17] , Csikai and 1325 1451 53 Chimoe et al. [47] were renormalized to the results of precise 1325 1451 54 absolute measurements of Ikeda et al.[53] in the overlapping ener-1325 1451 55 gy ranges. Correction factors for the experimental data [3], [8], 1325 1451 56 [15], [17] were Fc=0.87555, Fc=1.41467, Fc=1.60840, Fc=0.89474, 1325 1451 57 respectively. Data of independent measurements of Csikai and 1325 1451 58 Chimoe et al. [47] were multiplied to the coefficients Fc=0.89370 1325 1451 59 and Fc=0.92320, respectively. 1325 1451 60 Cross sections for the Al-27(n,p)Mg-27 reaction measured by 1325 1451 61 Bass et al. [12] in the neutron energy range 6.00 - 9.00 MeV with 1325 1451 62 25 keV step were recalculated by averaging original experimental 1325 1451 63 data over 100 keV energy intervals. 1325 1451 64 Experimental data of Calvi et al. [13] and Shimizu et al.[58] 1325 1451 65 were corrected to the results of Smith and Meadows measurements 1325 1451 66 with Li-7(p,n)Be-7 neutron source [27] in the overlapping energy 1325 1451 67 intervals. Correction factors were Fc=0.96506 and Fc=1.47768, res-1325 1451 68 pectively. For the experimental data [58] value Fc=1.47768 is a 1325 1451 69 total correction factor. On the first step of correction data of 1325 1451 70 Shimizu et al. [58] were renormalized to the new evaluated cross 1325 1451 71 sections for the monitor reaction In-115(n,n')In-115m [78]. 1325 1451 72 Data of Smith and Meadows [27] measured with using neutrons 1325 1451 73 from D(d,n)He3 reaction were renormalized to the results of this 1325 1451 74 experiment obtained with Li7(p,n)Be7 neutron source in the over- 1325 1451 75 lapping interval 5.398 - 5.870 MeV. D(d,n)He3 data in the energy 1325 1451 76 range 5.398 - 9.897 MeV were increased to the factor Fc=1.083 . 1325 1451 77 Data given in ref. [28] by Mostafa were renormalized to the 1325 1451 78 absolute cross section value for Al-27(n,p)Mg-27 reaction evalu- 1325 1451 79 ated at 7.1 MeV with taking into account experimental data [12], 1325 1451 80 [14] and [27]. 1325 1451 81 Experimental data [5], [8], [12] and [44] were used partially.1325 1451 82 Data of Mani et al. [5] were taken into account only for the 9 1325 1451 83 neutron energies: 11.92, 13.82, 14.75, 16.20, 16.63, 17.83, 18.13,1325 1451 84 20.15 and 21.72 MeV. Renormalized experimental data of Gabbard et 1325 1451 85 al. [8] were used only for the 8 neutron energies: 12.90, 13.10,1325 1451 86 13.70, 14.40, 14.90, 15.45, 16.80 and 17.40 MeV. Information ob- 1325 1451 87 tained from Bass et al. experiment [12] was taken into account in 1325 1451 88 the evaluation in the energy region 6.0 - 8.6 MeV. In the energy 1325 1451 89 interval 8.6 - 9.0 MeV this experiment gives significantly over- 1325 1451 90 estimated cross sections in a comparison with corrected experimen-1325 1451 91 tal data [24], [27] and new experimental data [55]. Cross section 1325 1451 92 measured by Bradley et al. [44] for the neutron energy 11 MeV was 1325 1451 93 not input in the data base for evaluation due to a very big dis- 1325 1451 94 crepancy with new experimental data of Csikai et al. [55]. 1325 1451 95 Experimental cross section data [59-76] were rejected due to 1325 1451 96 their discrepancy with the main bulk of experimental data [1-58]. 1325 1451 97 In the rejected experiments [59-70] and [75-76] cross section 1325 1451 98 values were measured only in a one energy point in the interval 1325 1451 99 14 - 15 MeV. 1325 1451 100 Statistical analysis of input cross section data was carried 1325 1451 101 out by means of PADE-2 code [79]. Rational function was used as 1325 1451 102 the model function [80]. 1325 1451 103 Evaluated excitation function for the reaction Al27(n,p)Mg27 1325 1451 104 was tested with using integral experimental data [81-83] for 1325 1451 105 U-235 thermal fission neutron spectrum and evaluated integral ex- 1325 1451 106 perimental data [83] for Cf-252 spontaneous fission neutron spec- 1325 1451 107 trum. Calculated and measured average cross section values for 1325 1451 108 U-235 thermal fission neutron spectrum [84] and Cf-252 sponta- 1325 1451 109 neous fission neutron spectrum [85] are given in the table 1. 1325 1451 110 Table 1 1325 1451 111 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 1325 1451 112 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 1325 1451 113 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 1325 1451 114 U-235 neutron fission ³ 4.0768 ³ 4.133 +- 0.074 [81] 1325 1451 115 ³ ³ 3.914 +- 0.070 [82] 1325 1451 116 ³ ³ 3.902 +- 0.069 [83] 1325 1451 117 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 1325 1451 118 CF-252 spont. fission ³ 4.9070 ³ 4.880 +- 0.104 [83] 1325 1451 119 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 1325 1451 120 1325 1451 121 MF=33 1325 1451 122 MT=103 - (n,p) cross section cov. matrix 1325 1451 123 ---------------------------------------- 1325 1451 124 Uncertainties in the evaluated excitation function for the 1325 1451 125 reaction Al-27(n,p)Mg-27 are given in the form of relative covari-1325 1451 126 ance matrix for the 49-neutron energy groups (LB=5). Covariance 1325 1451 127 matrix of uncertainties was calculated simultaneously with 1325 1451 128 recommended cross section data by means of PADE-2 code [79]. 1325 1451 129 Eigenvalues of the 6-th digits relative covariance matrix 1325 1451 130 given in the 33-file are the following: 1325 1451 131 1325 1451 132 9.62022E-08 9.94230E-08 1.04190E-07 1.09967E-07 1325 1451 133 1.15699E-07 1.20174E-07 1.24951E-07 1.30570E-07 1325 1451 134 1.37068E-07 1.43716E-07 1.50173E-07 1.58001E-07 1325 1451 135 1.68205E-07 1.76372E-07 1.86739E-07 1.94590E-07 1325 1451 136 2.06816E-07 2.14527E-07 2.29525E-07 2.40453E-07 1325 1451 137 2.53581E-07 2.82127E-07 2.99422E-07 3.49389E-07 1325 1451 138 3.68505E-07 4.35939E-07 5.42544E-07 7.20462E-07 1325 1451 139 9.45090E-07 2.44284E-06 1.77944E-05 8.88268E-05 1325 1451 140 1.54306E-04 2.21515E-04 2.92421E-04 3.30954E-04 1325 1451 141 3.45302E-04 3.84750E-04 4.56266E-04 6.07458E-04 1325 1451 142 1.31442E-03 1.53210E-03 1.76053E-03 2.87076E-03 1325 1451 143 4.69639E-03 5.64972E-03 1.04479E-02 1.48842E-02 1325 1451 144 5.19587E-02 1325 1451 145 1325 1451 146 References : 1325 1451 147 1. S.G.Forbes Phys. Rev., v.88, p.1309, December 1952 1325 1451 148 2. R.L.Henkel EXFOR 11524.002, 1954 1325 1451 149 3. O.M.Hudson jr, I.L.Morgan Bull. Am. Phys. Soc., v.4, p.97 1325 1451 150 (G2), March 1959 1325 1451 151 4. M.J.Depraz et al. Journal de Physique-Colloque, v.21, p.377, 1325 1451 152 May 1960 1325 1451 153 5. G.S.Mani et.al. Nucl. Phys., v.19, n.5, p.535, November 1960 1325 1451 154 6. H.Pollehn, H.Neuert Zeitschrift fuer Naturforschung, Sect. A,1325 1451 155 v.16, p.227, 1961 1325 1451 156 7. S.K.Mukherjee et al. Proc. of the Physical Society, v.77, 1325 1451 157 p.508, February 1961 1325 1451 158 8. F.Gabbard, B.D.Kern Phys. Rev., v.128, p.1276, 1962 1325 1451 159 9. J.Csikai et al. Atomki Koezlemenyek v.4, p.137, June 1962 ; 1325 1451 160 J.Csikai et al. Nucl. Phys., v.46, p.141, July 1963 1325 1451 161 10. C.G.Bonazzola et al. Nucl. Phys., v.51, p.337, February 1964 1325 1451 162 11. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 1325 1451 163 12. R.Bass et al. 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H.A.Husain, S.E.Hunt International J. of Applied Radiation 1325 1451 208 and Isotopes, v.34, no.4, p.731, 1983 1325 1451 209 43. V.T.Shchebolev et al. Atomnaya Energiya, v.54, no.6, p.417, 1325 1451 210 June 1983 1325 1451 211 44. D.A.Bradley et al. Proc. of Int. Symp. on Fast Neutrons in 1325 1451 212 Science and Technology, Chiang Mai, 4-8 February 1985, p.19 1325 1451 213 45. W.Enz et al. Annalen der Physik, v.42, no.3, p.283, 1985 1325 1451 214 46. I.Garlea et al. Rev. Roum. Phys., v.30, no.8, p.673, 1985 1325 1451 215 47. J.Csikai, T.Chimoye et al. Zeitschrift fuer Physik, Sec.A, 1325 1451 216 v.325, p.69, September 1986 1325 1451 217 48. J.W.Meadows et al. Ann. Nucl. Energ., v.14, p.489, Sep. 1987 1325 1451 218 49. Y.Ikeda,C.Konno,K.Oishi et al. Report JAERI-1312, March 1988 1325 1451 219 50. K.Kudo et.al. Proc. of International Conference Nuclear Data 1325 1451 220 for Science and Technology, Mito, Japan, 30 May - 3 June 1988,1325 1451 221 Saikon Publishing Co., LTD, pp.1021-1024, 1989 1325 1451 222 51. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 ; 1325 1451 223 K.Kobayashi, I.Kimura Proc. of the Intern. Conf. on Nuclear 1325 1451 224 Data for Science and Technology, 30 May - 3 June 1988, Mito, 1325 1451 225 Japan, Saikon Publishing Co., LTD, pp.261-265, 1989 ; 1325 1451 226 K.Kobayashi, I.Kimura Progress Report NEANDC(J)-116, Sep.1985 1325 1451 227 52. A.Ercan et al. Proc. of an Intern. Conf. on Nuclear Data for 1325 1451 228 Science and Technology, 13-17 May 1991, Julich, FRG, Springer-1325 1451 229 Verlag, 1992 1325 1451 230 53. Y.Ikeda et al. Journal of Nuclear Science and Technology, 1325 1451 231 v.30, n.9, pp.870-880, September 1993 1325 1451 232 54. Zhou Hongyu et al. Proc. of Int. Conf. on Nuclear Data for 1325 1451 233 Science and Technology, Gatlinburg, Tennessee, USA, May 9-13, 1325 1451 234 1994, Vol.1, pp. 166-169 1325 1451 235 55. J.Csikai et al. Private communication, Debrecen, April 1998 1325 1451 236 56. A.A.Filatenkov et al. Report RI-252, St.Petersburg, May 1999; 1325 1451 237 A.A.Filatenkov et al. VANT, Ser.:Yadernye Konstanty, v.2, p.8,1325 1451 238 Moscow, 1996 1325 1451 239 57. A.Fessler, A.J.M.Plompen et al. Nucl. Sci. Eng., v.134, no.2, 1325 1451 240 pp.171-200, February 2000 1325 1451 241 58. T.Shimizu, S.Furuichi, H.Sakane, M.Shibata, K.Kawade 1325 1451 242 Proc. of The 2000 Symposium on Nuclear Data, November 16-17, 1325 1451 243 2000 JAERI, Tokai, Japan, pp.194-199 1325 1451 244 59. E.B.Paul, R.L.Clarke Canadian Journal of Physics, v.31, 1325 1451 245 p.267, 1953 1325 1451 246 60. S.Yasumi Journal of the Physical Society of Japan, v.12, 1325 1451 247 p.443, May 1957 1325 1451 248 61. G.Brown et al. Philosophical Magazine, v.2, p.785, 1957 1325 1451 249 62. A.Poularikas, R.W.Fink Phys. Rev., v.115, p.989, 1959 1325 1451 250 63. C.S.Khurana, H.S.Hans Proc. of 4th Nuclear Physics and Solid 1325 1451 251 Stata Physics Symp., 24-26 Feb. 1960, Waltair, India, p.297 1325 1451 252 64. R.S.Storey,W.Jack,A.Ward Proc. Phys. Soc., v.75, p.526, 1960 1325 1451 253 65. M.Sakisaka et al. Journal of the Physical Society of Japan, 1325 1451 254 v.16, p.1869, October 1961 1325 1451 255 66. J.Kantele, D.G.Gardner Nucl. Phys, v.35, p.353, 1962 1325 1451 256 67. W.Langmann EXFOR 20903.003, September 1962 1325 1451 257 68. F.L.Hassler, R.A.Peck jr Phys. Rev., v.125, p.1011, 1962 1325 1451 258 69. B.Mitra, A.M.Ghose Nucl. Phys., v.83, p.157, July 1966 1325 1451 259 70. R.Prasad, D.C.Sarkar Nuovo Cimento, v.A3, no.3, p.467, 1971 1325 1451 260 71. J.C.Robertson, K.J.Zieba Annals of Nucl. Energy, v.26, no.1, 1325 1451 261 p.1, 1972 1325 1451 262 72. R.A.Jarjis J. of Physics, pt.G, v.4, n.3, p.445, 1978 1325 1451 263 73. J.Csikai Proc. of International Conference Nuclear Data for 1325 1451 264 Science and Technology, 6 - 10 September 1982, Antwerp, 1325 1451 265 Holland, D.Reidel Publishing Company, p.414, 1983 1325 1451 266 74. Tahir Indian Journal of Pure and Applied Physics, v.23, 1325 1451 267 p.439, September 1985 1325 1451 268 75. J.P.Gupta et.al. Indian J. Pramana, v.24, p.637, 1985 1325 1451 269 76. L.I.Klochkova et al. Vopr. Atomn. Nauki i Tekhn., Serija: 1325 1451 270 Jadernye Konstanty v.1, p.27, 1992 ; 1325 1451 271 L.I.Klochkova et al. Proc. of the 1-st Int. Conf. on Neutron 1325 1451 272 Phys., Kiev, USSR, 14 - 18 Sep.1987, v.3, p.315, Moscow 1988 1325 1451 273 77. R.B.Firestone Table of Isotopes, Eighth edition, Vol. 1, 1325 1451 274 John Wiley & Sons, Inc., New York, 1995 1325 1451 275 78. K.I.Zolotarev, P.K.Zolotarev The In-115(n,n')In-115m reaction 1325 1451 276 excitation function in the energy range from threshold to 1325 1451 277 20 MeV, Obninsk, IPPE, eval. March 2003 1325 1451 278 79. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 1325 1451 279 80. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 1325 1451 280 st's Meeting on Evaluation and Processing of Covariance Data, 1325 1451 281 Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105 1325 1451 282 81. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 1325 1451 283 Washington D.C., 25-28 April 1989, v.2, p.923 1325 1451 284 82. W.Mannhart Prog. Report INDC(Ger)-045, pp.40-43, June 1999 1325 1451 285 83. W.Mannhart Validation of Differential Cross Sections with 1325 1451 286 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 1325 1451 287 September 2002 1325 1451 288 84. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 1325 1451 289 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 1325 1451 290 85. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 1325 1451 291 ***************************************************************** 1325 1451 292 ********** End of (N,2N) bibliographical component ********** 1325 1451 293 ================================================================= 1325 1451 294 ***************************************************************** 1325 1451 295 The Q values and threshold energies were updated prior to pro- 1325 1451 296 cessing through the codes to comply with the values obtained 1325 1451 297 using the NNDC calculation program which is based on the 1995 1325 1451 298 Update to the Atomic mass Evaluation.Additional points were 1325 1451 299 added near threshold for MF/MT=3/107 1325 1451 300 File 2 added to the pointwise file containing only the effective 1325 1451 301 scattering radius with no resonance parameters given. 1325 1451 302 Taken from ENDF/B-VI 1325 1451 303 ***************************************************************** 1325 1451 304 ***************** Program LINEAR (VERSION 2002-1) ***************1325 1451 305 For All Data Greater than 1.0000E-10 barns in Absolute Value 1325 1451 306 Data Linearized to Within an Accuracy of .100000000 per-cent 1325 1451 307 ***************** Program SIGMA1 (VERSION 2002-1) ***************1325 1451 308 Data Doppler Broadened to 300.000000 Kelvin 1325 1451 309 for All Data Greater than 1.0000E-10 barns in Absolute Value 1325 1451 310 Data Linearized to Within an Accuracy pf .100000000 per-cent 1325 1451 311 ***************** Program FIXUP (Version 2002-1) ****************1325 1451 312 Corrected ZA/AWR in All Sections-----------------------------Yes 1325 1451 313 Corrected Thresholds-----------------------------------------Yes 1325 1451 314 Extended Cross Sections to 20 MeV----------------------------No 1325 1451 315 Allow Cross Section Deletion---------------------------------No 1325 1451 316 Allow Cross Section Reconstruction---------------------------No 1325 1451 317 Make All Cross Sections Non-Negative-------------------------Yes 1325 1451 318 Delete Energies Not in Ascending Order-----------------------Yes 1325 1451 319 Deleted Duplicate Points-------------------------------------Yes 1325 1451 320 Check for Ascending MAT/MF/MT Order--------------------------Yes 1325 1451 321 Check for Legal MF/MT Numbers--------------------------------Yes 1325 1451 322 Allow Creation of Missing Sections---------------------------No 1325 1451 323 Allow Insertion of Energy Points-----------------------------No 1325 1451 324 Create Uniform Energy Grid-----------------------------------No 1325 1451 325 Delete Section if Cross Section =0 at All Energies-----------Yes 1325 1451 326 ***************** Program GROUPIE (VERSION 2002-1) **************1325 1451 327 Unshielded Group Averages Using 640 Groups 1325 1451 328 Weighting Spectrum: Flat (Constant) Spectrum 1325 1451 329 1 451 335 11325 1451 330 2 151 4 11325 1451 331 3 103 64 11325 1451 332 3 107 60 11325 1451 333 33 103 224 11325 1451 334 33 107 224 11325 1451 335 1325 1 099999 1325 0 0 0 1.30270E+4 2.67497E+1 0 0 1 01325 2151 1 1.302700+4 1.000000+0 0 0 1 01325 2151 2 1.000000-5 1.000000+3 0 0 0 01325 2151 3 2.500000+0 3.275200-1 0 0 0 01325 2151 4 1325 2 099999 1325 0 0 0 1.30270E+4 2.67497E+1 0 0 0 01325 3103 1 -1.82797E+6-1.82797E+6 0 0 1 1831325 3103 2 183 1 1325 3103 3 1800000.00 7.5592E-10 1900000.00 5.94923E-7 2000000.00 1.70291E-61325 3103 4 2100000.00 2.81090E-6 2200000.00 3.91889E-6 2300000.00 5.02688E-61325 3103 5 2400000.00 1.36533E-5 2500000.00 5.31887E-5 2600000.00 9.60656E-51325 3103 6 2700000.00 .000197747 2800000.00 .000422101 2900000.00 .0007430851325 3103 7 3000000.00 .001163323 3100000.00 .001664055 3200000.00 .0021825181325 3103 8 3300000.00 .002660038 3400000.00 .005813040 3500000.00 .0059058401325 3103 9 3600000.00 .005923953 3700000.00 .007376778 3800000.00 .0053420301325 3103 10 3900000.00 .006590330 4000000.00 .006400630 4100000.00 .0058240901325 3103 11 4200000.00 .010016250 4300000.00 .011382425 4400000.00 .0120621751325 3103 12 4500000.00 .017318350 4600000.00 .018967950 4700000.00 .0161955251325 3103 13 4800000.00 .015435425 4900000.00 .023968600 5000000.00 .0268780501325 3103 14 5100000.00 .028133700 5200000.00 .028140700 5300000.00 .0276018001325 3103 15 5400000.00 .036424900 5500000.00 .041505550 5600000.00 .0478750101325 3103 16 5700000.00 .058418765 5800000.00 .042922684 5900000.00 .0452353451325 3103 17 6000000.00 .048209582 6100000.00 .051044500 6200000.00 .0536079001325 3103 18 6300000.00 .055899200 6400000.00 .057914600 6500000.00 .0596261001325 3103 19 6600000.00 .061243488 6700000.00 .062934663 6800000.00 .0646258381325 3103 20 6900000.00 .066317013 7000000.00 .067955880 7100000.00 .0695424401325 3103 21 7200000.00 .071129000 7300000.00 .072715560 7400000.00 .0743021201325 3103 22 7500000.00 .075839390 7600000.00 .077327370 7700000.00 .0788153501325 3103 23 7800000.00 .080303330 7900000.00 .081791310 8000000.00 .0832272701325 3103 24 8100000.00 .084611210 8200000.00 .085995150 8300000.00 .0873790901325 3103 25 8400000.00 .088763030 8500000.00 .090087063 8600000.00 .0913511881325 3103 26 8700000.00 .092615313 8800000.00 .093879438 8900000.00 .0950713751325 3103 27 9000000.00 .096191125 9100000.00 .097310875 9200000.00 .0984306251325 3103 28 9300000.00 .099468917 9400000.00 .100425750 9500000.00 .1013825831325 3103 29 9600000.00 .102253820 9700000.00 .103039461 9800000.00 .1038251011325 3103 30 9900000.00 .104511118 10000000.0 .105097471 10100000.0 .1056838241325 3103 31 10200000.0 .106158000 10300000.0 .106520000 10400000.0 .1068820001325 3103 32 10500000.0 .107122667 10600000.0 .107242000 10700000.0 .1073613331325 3103 33 10800000.0 .107354500 10900000.0 .107221500 11000000.0 .1070885001325 3103 34 11100000.0 .106830333 11200000.0 .106447000 11300000.0 .1060636671325 3103 35 11400000.0 .105561500 11500000.0 .104940500 11600000.0 .1043195001325 3103 36 11700000.0 .103591167 11800000.0 .102755500 11900000.0 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1.952000-5 3.840000-7 1.564000-5 1.574000-5 0.000000+0132533107 213 1.663000-5 5.067000-6 6.248000-6 1.517000-6 1.349000-6 1.744000-6132533107 214 2.882000-6 1.078000-5 1.169000-5 1.572000-5 1.178000-5 1.555000-5132533107 215 5.900000-6 7.406000-6 1.009000-5 4.208000-6 1.284000-5 3.768000-6132533107 216 1.077000-5 1.148000-5 1.088000-5 7.103000-5 4.890000-6 1.567000-6132533107 217 2.252000-6 3.467000-6 1.148000-6 1.616000-6 1.783000-6 8.398000-6132533107 218 1.958000-4 1.062000-4 1.967000-5 1.202000-4 2.103000-5 1.000000-4132533107 219 1.130000-4 0.000000+0 3.040000-4 3.899000-5 1.845000-4 3.677000-5132533107 220 1.565000-4 1.772000-4 1.813000-5 4.736000-4 3.914000-5 5.423000-5132533107 221 5.088000-5 7.374000-5 2.934000-5 4.221000-4 7.744000-5 2.208000-4132533107 222 2.874000-4 2.662000-4 4.119000-4 1.790000-4 2.601000-4 1.574000-4132533107 223 5.736000-4 4.115000-4 2.375000-4 1.008000-3 3.285000-4 2.638000-3132533107 224 132533 099999 1325 0 0 0 0 0 0 0 1.50310E+4 3.07076E+1 0 0 34 101525 1451 1 0.0 0.0 0 0 0 61525 1451 2 1.00000E+0 2.05000E+7 0 0 10 20021525 1451 3 3.00000E+2 0.0 1 0 40 41525 1451 4 15-P - 31 IRK-VIENNA EVAL-JUN80 1525 1451 5 DIST-Feb2004 1525 1451 6 ----IRDF-2002 MATERIAL 1525 1525 1451 7 -----INCIDENT NEUTRON DATA 1525 1451 8 ------ENDF-6 FORMAT 1525 1451 9 *****************************************************************1525 1451 10 DIST-JUN90 1525 1451 11 IRK-EVAL.NLIB 25 1525 1525 1451 12 *****************************************************************1525 1451 13 The Q values and threshold energies were updated prior to pro- 1525 1451 14 cessing through the codes to comply with the values obtained 1525 1451 15 using the NNDC calculation program which is based on the 1995 1525 1451 16 Update to the Atomic mass Evaluation. 1525 1451 17 Additional points were added near threshold for mf/mt= 3/103 1525 1451 18 *****************************************************************1525 1451 19 ***************** Program LINEAR (VERSION 2002-1) ***************1525 1451 20 For All Data Greater than 1.0000E-10 barns in Absolute Value 1525 1451 21 Data Linearized to Within an Accuracy of .100000000 per-cent 1525 1451 22 ***************** Program SIGMA1 (VERSION 2002-1) ***************1525 1451 23 Data Doppler Broadened to 300.000000 Kelvin 1525 1451 24 for All Data Greater than 1.0000E-10 barns in Absolute Value 1525 1451 25 Data Linearized to Within an Accuracy pf .100000000 per-cent 1525 1451 26 ***************** Program FIXUP (Version 2002-1) ****************1525 1451 27 Corrected ZA/AWR in All Sections-----------------------------Yes 1525 1451 28 Corrected Thresholds-----------------------------------------Yes 1525 1451 29 Extended Cross Sections to 20 MeV----------------------------No 1525 1451 30 Allow Cross Section Deletion---------------------------------No 1525 1451 31 Allow Cross Section Reconstruction---------------------------No 1525 1451 32 Make All Cross Sections Non-Negative-------------------------Yes 1525 1451 33 Delete Energies Not in Ascending Order-----------------------Yes 1525 1451 34 Deleted Duplicate Points-------------------------------------Yes 1525 1451 35 Check for Ascending MAT/MF/MT Order--------------------------Yes 1525 1451 36 Check for Legal MF/MT Numbers--------------------------------Yes 1525 1451 37 Allow Creation of Missing Sections---------------------------No 1525 1451 38 Allow Insertion of Energy Points-----------------------------No 1525 1451 39 Create Uniform Energy Grid-----------------------------------No 1525 1451 40 Delete Section if Cross Section =0 at All Energies-----------Yes 1525 1451 41 ***************** Program GROUPIE (VERSION 2002-1) **************1525 1451 42 Unshielded Group Averages Using 640 Groups 1525 1451 43 Weighting Spectrum: Flat (Constant) Spectrum 1525 1451 44 1 451 48 11525 1451 45 2 151 4 11525 1451 46 3 103 69 11525 1451 47 33 103 108 11525 1451 48 1525 1 099999 1525 0 0 0 1.50310E+4 3.07076E+1 0 0 1 01525 2151 1 1.50310E+4 1.00000E+0 0 0 1 01525 2151 2 1.00000E+0 2.00000E+7 0 0 0 01525 2151 3 5.00000E-1 4.10000E-1 0 0 0 01525 2151 4 1525 2 099999 1525 0 0 0 1.50310E+4 3.07076E+1 0 0 0 01525 3103 1 -7.09680E+5-7.09680E+5 0 0 1 1981525 3103 2 198 1 1525 3103 3 720000.000 4.99267E-7 760000.000 2.54695E-6 800000.000 4.70496E-61525 3103 4 840000.000 6.86298E-6 880000.000 9.02099E-6 920000.000 1.43222E-51525 3103 5 960000.000 2.27667E-5 1000000.00 3.75444E-5 1100000.00 8.10667E-51525 3103 6 1200000.00 .000175412 1300000.00 .000440231 1400000.00 .0007334621525 3103 7 1500000.00 .001126279 1600000.00 .003710000 1700000.00 .0064587501525 3103 8 1800000.00 .007420000 1900000.00 .008352500 2000000.00 .0093375001525 3103 9 2100000.00 .014567500 2200000.00 .024042500 2300000.00 .0318600001525 3103 10 2400000.00 .038020000 2500000.00 .046500000 2600000.00 .0573000001525 3103 11 2700000.00 .062522500 2800000.00 .062167500 2900000.00 .0614850001525 3103 12 3000000.00 .060475000 3100000.00 .063037500 3200000.00 .0691725001525 3103 13 3300000.00 .070750000 3400000.00 .067770000 3500000.00 .0675950001525 3103 14 3600000.00 .070225000 3700000.00 .075490000 3800000.00 .0833900001525 3103 15 3900000.00 .086990000 4000000.00 .086290000 4100000.00 .0890725001525 3103 16 4200000.00 .095337500 4300000.00 .099325000 4400000.00 .1010350001525 3103 17 4500000.00 .106707500 4600000.00 .116342500 4700000.00 .1218066671525 3103 18 4800000.00 .123100000 4900000.00 .124393333 5000000.00 .1252008331525 3103 19 5100000.00 .125522500 5200000.00 .125844167 5300000.00 .1261658331525 3103 20 5400000.00 .126487500 5500000.00 .126809167 5600000.00 .1277855561525 3103 21 5700000.00 .129416667 5800000.00 .131047778 5900000.00 .1326788891525 3103 22 6000000.00 .134310000 6100000.00 .135941111 6200000.00 .1375722221525 3103 23 6300000.00 .139203333 6400000.00 .140834444 6500000.00 .1411960001525 3103 24 6600000.00 .140288000 6700000.00 .139380000 6800000.00 .1384720001525 3103 25 6900000.00 .137564000 7000000.00 .136656000 7100000.00 .1357480001525 3103 26 7200000.00 .134840000 7300000.00 .133932000 7400000.00 .1330240001525 3103 27 7500000.00 .132823600 7600000.00 .133330800 7700000.00 .1338380001525 3103 28 7800000.00 .134345200 7900000.00 .134852400 8000000.00 .1353596001525 3103 29 8100000.00 .135866800 8200000.00 .136374000 8300000.00 .1368812001525 3103 30 8400000.00 .137388400 8500000.00 .137895600 8600000.00 .1384028001525 3103 31 8700000.00 .138892500 8800000.00 .139277200 8900000.00 .1396444001525 3103 32 9000000.00 .140011600 9100000.00 .140378800 9200000.00 .1407460001525 3103 33 9300000.00 .141113200 9400000.00 .141480400 9500000.00 .1418476001525 3103 34 9600000.00 .142214800 9700000.00 .142582000 9800000.00 .1429492001525 3103 35 9900000.00 .143316400 10000000.0 .143224500 10100000.0 .1426735001525 3103 36 10200000.0 .142122500 10300000.0 .141571500 10400000.0 .1410205001525 3103 37 10500000.0 .140469500 10600000.0 .139918500 10700000.0 .1393675001525 3103 38 10800000.0 .138816500 10900000.0 .138265500 11000000.0 .1371955001525 3103 39 11100000.0 .135606500 11200000.0 .134017500 11300000.0 .1324285001525 3103 40 11400000.0 .130839500 11500000.0 .129250500 11600000.0 .1276615001525 3103 41 11700000.0 .126072500 11800000.0 .124483500 11900000.0 .1228945001525 3103 42 12000000.0 .121343750 12100000.0 .119831250 12200000.0 .1183187501525 3103 43 12300000.0 .116806250 12400000.0 .115293750 12500000.0 .1137812501525 3103 44 12600000.0 .112268750 12700000.0 .110756250 12800000.0 .1092437501525 3103 45 12900000.0 .107731250 13000000.0 .106218750 13100000.0 .1047062501525 3103 46 13200000.0 .103193750 13300000.0 .101681250 13400000.0 .1001687501525 3103 47 13500000.0 .098656250 13600000.0 .097143750 13700000.0 .0956312501525 3103 48 13800000.0 .094118750 13900000.0 .092606250 14000000.0 .0914250001525 3103 49 14100000.0 .090575000 14200000.0 .089725000 14300000.0 .0888750001525 3103 50 14400000.0 .088025000 14500000.0 .087175000 14600000.0 .0863250001525 3103 51 14700000.0 .085475000 14800000.0 .084625000 14900000.0 .0837750001525 3103 52 15000000.0 .082900500 15100000.0 .082001500 15200000.0 .0811025001525 3103 53 15300000.0 .080203500 15400000.0 .079304500 15500000.0 .0784055001525 3103 54 15600000.0 .077506500 15700000.0 .076607500 15800000.0 .0757085001525 3103 55 15900000.0 .074809500 16000000.0 .073979000 16100000.0 .0732170001525 3103 56 16200000.0 .072455000 16300000.0 .071693000 16400000.0 .0709310001525 3103 57 16500000.0 .070169000 16600000.0 .069407000 16700000.0 .0686450001525 3103 58 16800000.0 .067883000 16900000.0 .067121000 17000000.0 .0664115001525 3103 59 17100000.0 .065754500 17200000.0 .065097500 17300000.0 .0644405001525 3103 60 17400000.0 .063783500 17500000.0 .063126500 17600000.0 .0624695001525 3103 61 17700000.0 .061812500 17800000.0 .061155500 17900000.0 .0604985001525 3103 62 18000000.0 .059888500 18100000.0 .059325500 18200000.0 .0587625001525 3103 63 18300000.0 .058199500 18400000.0 .057636500 18500000.0 .0570735001525 3103 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0.000000+0152533103 73 3.230000-3 8.735000-5 5.405000-4 8.327000-5 1.096000-3 8.539000-4152533103 74 1.304000-3 1.275000-3 9.864999-4 9.834000-4 0.000000+0 0.000000+0152533103 75 0.000000+0 4.475000-4 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103 76 0.000000+0 0.000000+0 3.690000-3 2.769000-4 3.660000-4 9.568000-4152533103 77 9.384000-4 9.591000-4 9.271000-4 7.527000-4 6.426000-4 0.000000+0152533103 78 0.000000+0 0.000000+0 2.754000-4 0.000000+0 0.000000+0 0.000000+0152533103 79 0.000000+0 0.000000+0 0.000000+0 1.786000-3 2.632000-4 7.904000-4152533103 80 6.972000-4 7.748000-4 7.579000-4 5.864000-4 5.845000-4 0.000000+0152533103 81 0.000000+0 0.000000+0 2.660000-4 0.000000+0 0.000000+0 0.000000+0152533103 82 0.000000+0 0.000000+0 0.000000+0 2.754000-3 2.846000-4 2.387000-4152533103 83 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103 84 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103 85 0.000000+0 0.000000+0 5.841000-3 1.996000-3 2.692000-3 2.620000-3152533103 86 2.069000-3 1.934000-3 0.000000+0 0.000000+0 0.000000+0 8.599000-4152533103 87 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103 88 3.286000-3 2.544000-3 2.474000-3 1.958000-3 1.818000-3 0.000000+0152533103 89 0.000000+0 0.000000+0 8.061000-4 0.000000+0 0.000000+0 0.000000+0152533103 90 0.000000+0 0.000000+0 0.000000+0 4.903000-3 3.626000-3 2.867000-3152533103 91 2.670000-3 0.000000+0 0.000000+0 0.000000+0 1.186000-3 0.000000+0152533103 92 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 4.647000-3152533103 93 2.789000-3 2.599000-3 0.000000+0 0.000000+0 0.000000+0 1.154000-3152533103 94 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0152533103 95 3.739000-3 3.378000-3 4.381000-3 4.071000-3 3.459000-3 1.514000-3152533103 96 1.984000-4-1.096000-3-2.519000-3-3.665000-3-4.531000-3-5.258000-3152533103 97 3.602000-3 5.156000-3 4.879000-3 4.114000-3 1.819000-3 2.486000-4152533103 98 -1.353000-3-3.211000-3-4.813000-3-6.048000-3-7.052000-3 8.150000-3152533103 99 7.875000-3 6.715000-3 3.008000-3 4.616000-4-2.013000-3-4.899000-3152533103 100 -7.395000-3-9.318000-3-1.087000-2 7.667000-3 6.600000-3 3.022000-3152533103 101 5.465000-4-1.729000-3-4.440000-3-6.795000-3-8.611000-3-1.007000-2152533103 102 5.775000-3 2.746000-3 6.217000-4-1.125000-3-3.301000-3-5.194000-3152533103 103 -6.656000-3-7.841000-3 1.439000-3 4.791000-4-7.138000-5-9.163000-4152533103 104 -1.679000-3-2.276000-3-2.766000-3 3.212000-4 5.254000-4 5.616000-4152533103 105 5.582000-4 5.450000-4 5.245000-4 1.899000-3 3.081000-3 4.045000-3152533103 106 4.771000-3 5.329000-3 5.515000-3 7.581000-3 9.161000-3 1.039000-2152533103 107 1.064000-2 1.300000-2 1.484000-2 1.597000-2 1.828000-2 2.098000-2152533103 108 152533 099999 1525 0 0 0 0 0 0 0 1.60320E+4 3.16974E+1 0 0 34 101625 1451 1 0.0 0.0 0 0 0 61625 1451 2 1.00000E+0 2.00000E+7 0 0 10 20021625 1451 3 3.00000E+2 0.0 1 0 45 41625 1451 4 16-S - 32 1625 1451 5 DIST-Feb2004 1625 1451 6 ----IRDF-2002 MATERIAL 1625 1625 1451 7 -----INCIDENT NEUTRON DATA 1625 1451 8 ------ENDF-6 FORMAT 1625 1451 9 ******************************************************************1625 1451 10 US EVALUATION 1989 DIST-FEB91 1625 1451 11 ENDF/B-VI MATERIAL 1625 1625 1451 12 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************1625 1451 13 DOSIMETRY 1625 1451 14 ******************************************************************1625 1451 15 The Q values and threshold energies were updated prior to pro- 1625 1451 16 cessing through the codes to comply with the values obtained 1625 1451 17 using the NNDC calculation program which is based on the 1995 1625 1451 18 Update to the Atomic mass Evaluation. 1625 1451 19 1625 1451 20 File 2 added to the pointwise file containing only the effective 1625 1451 21 scattering radius with no resonance parameters given. 1625 1451 22 Taken from ENDF/B-VI 1625 1451 23 ******************************************************************1625 1451 24 ***************** Program LINEAR (VERSION 2002-1) ***************1625 1451 25 For All Data Greater than 1.0000E-10 barns in Absolute Value 1625 1451 26 Data Linearized to Within an Accuracy of .100000000 per-cent 1625 1451 27 ***************** Program SIGMA1 (VERSION 2002-1) ***************1625 1451 28 Data Doppler Broadened to 300.000000 Kelvin 1625 1451 29 for All Data Greater than 1.0000E-10 barns in Absolute Value 1625 1451 30 Data Linearized to Within an Accuracy pf .100000000 per-cent 1625 1451 31 ***************** Program FIXUP (Version 2002-1) ****************1625 1451 32 Corrected ZA/AWR in All Sections-----------------------------Yes 1625 1451 33 Corrected Thresholds-----------------------------------------Yes 1625 1451 34 Extended Cross Sections to 20 MeV----------------------------No 1625 1451 35 Allow Cross Section Deletion---------------------------------No 1625 1451 36 Allow Cross Section Reconstruction---------------------------No 1625 1451 37 Make All Cross Sections Non-Negative-------------------------Yes 1625 1451 38 Delete Energies Not in Ascending Order-----------------------Yes 1625 1451 39 Deleted Duplicate Points-------------------------------------Yes 1625 1451 40 Check for Ascending MAT/MF/MT Order--------------------------Yes 1625 1451 41 Check for Legal MF/MT Numbers--------------------------------Yes 1625 1451 42 Allow Creation of Missing Sections---------------------------No 1625 1451 43 Allow Insertion of Energy Points-----------------------------No 1625 1451 44 Create Uniform Energy Grid-----------------------------------No 1625 1451 45 Delete Section if Cross Section =0 at All Energies-----------Yes 1625 1451 46 ***************** Program GROUPIE (VERSION 2002-1) **************1625 1451 47 Unshielded Group Averages Using 640 Groups 1625 1451 48 Weighting Spectrum: Flat (Constant) Spectrum 1625 1451 49 1 451 53 11625 1451 50 2 151 4 11625 1451 51 3 103 68 11625 1451 52 33 103 183 11625 1451 53 1625 1 099999 1625 0 0 0 1.60320E+4 3.16974E+1 0 0 1 01625 2151 1 1.603200+4 1.000000+0 0 0 1 01625 2151 2 1.000000-5 2.000000+7 0 0 0 01625 2151 3 0.000000+0 3.100000-1 0 0 0 01625 2151 4 1625 2 099999 1625 0 0 0 1.60320E+4 3.16974E+1 0 0 0 01625 3103 1 0.0 -9.28310E+5 0 0 1 1931625 3103 2 193 1 1625 3103 3 920000.000 2.17952E-8 960000.000 6.76274E-6 1000000.00 1.89000E-51625 3103 4 1100000.00 3.75000E-5 1200000.00 8.10000E-5 1300000.00 .0001710001625 3103 5 1400000.00 .000350000 1500000.00 .000688000 1600000.00 .0014590001625 3103 6 1700000.00 .002604000 1800000.00 .006027000 1900000.00 .0090950001625 3103 7 2000000.00 .013095000 2100000.00 .031680400 2200000.00 .0543095001625 3103 8 2300000.00 .068355500 2400000.00 .071605250 2500000.00 .0723635001625 3103 9 2600000.00 .066761000 2700000.00 .079601500 2800000.00 .0940746001625 3103 10 2900000.00 .112995500 3000000.00 .160350000 3100000.00 .1612500001625 3103 11 3200000.00 .147160000 3300000.00 .203350000 3400000.00 .2110525001625 3103 12 3500000.00 .227087500 3600000.00 .226475000 3700000.00 .1787000001625 3103 13 3800000.00 .178425000 3900000.00 .251806667 4000000.00 .3165733331625 3103 14 4100000.00 .340450000 4200000.00 .349000000 4300000.00 .3469636361625 3103 15 4400000.00 .296862864 4500000.00 .278430000 4600000.00 .2780800001625 3103 16 4700000.00 .272954286 4800000.00 .257940851 4900000.00 .2369763641625 3103 17 5000000.00 .220432500 5100000.00 .213717500 5200000.00 .2020425001625 3103 18 5300000.00 .185407500 5400000.00 .200845000 5500000.00 .2483550001625 3103 19 5600000.00 .278015000 5700000.00 .289825000 5800000.00 .2966925001625 3103 20 5900000.00 .298617500 6000000.00 .300240000 6100000.00 .3015600001625 3103 21 6200000.00 .302880000 6300000.00 .304200000 6400000.00 .3055200001625 3103 22 6500000.00 .307460000 6600000.00 .310020000 6700000.00 .3125800001625 3103 23 6800000.00 .315140000 6900000.00 .317700000 7000000.00 .3208110001625 3103 24 7100000.00 .324473000 7200000.00 .328135000 7300000.00 .3317970001625 3103 25 7400000.00 .335459000 7500000.00 .334650000 7600000.00 .3293700001625 3103 26 7700000.00 .324090000 7800000.00 .318810000 7900000.00 .3135300001625 3103 27 8000000.00 .312126000 8100000.00 .314598000 8200000.00 .3170700001625 3103 28 8300000.00 .319542000 8400000.00 .322014000 8500000.00 .3238300001625 3103 29 8600000.00 .324990000 8700000.00 .326150000 8800000.00 .3273100001625 3103 30 8900000.00 .328470000 9000000.00 .330531000 9100000.00 .3334930001625 3103 31 9200000.00 .336455000 9300000.00 .339417000 9400000.00 .3423790001625 3103 32 9500000.00 .345761000 9600000.00 .349563000 9700000.00 .3533650001625 3103 33 9800000.00 .357167000 9900000.00 .360969000 10000000.0 .3643230001625 3103 34 10100000.0 .367229000 10200000.0 .370135000 10300000.0 .3730410001625 3103 35 10400000.0 .375947000 10500000.0 .378436000 10600000.0 .3805080001625 3103 36 10700000.0 .382580000 10800000.0 .384652000 10900000.0 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2.4500E+06 7.8821E-06162533103 99 2.5000E+06 1.2628E-05 2.5500E+06 5.9731E-06 2.6000E+06 1.4956E-05162533103 100 2.6500E+06 5.6757E-06 2.7000E+06 3.0742E-05 2.7500E+06 2.1495E-05162533103 101 2.8000E+06 9.6057E-06 2.8400E+06 1.4095E-05 2.9000E+06 1.4186E-05162533103 102 2.9500E+06 2.7937E-05 3.0000E+06 4.9213E-05 3.0500E+06 9.8839E-05162533103 103 3.1000E+06 7.4618E-05 3.1500E+06 9.1742E-05 3.2000E+06 4.4272E-05162533103 104 3.2400E+06 1.2001E-04 3.2800E+06 9.5391E-05 3.3300E+06 1.4439E-04162533103 105 3.3700E+06 1.0161E-04 3.4000E+06 8.4846E-05 3.4400E+06 1.1310E-04162533103 106 3.4700E+06 1.3775E-04 3.5100E+06 1.0996E-04 3.5500E+06 5.8369E-05162533103 107 3.6000E+06 1.2237E-04 3.6500E+06 7.8251E-05 3.7000E+06 1.1350E-04162533103 108 3.7500E+06 1.4525E-04 3.8000E+06 8.7453E-05 3.8500E+06 1.7881E-04162533103 109 3.9000E+06 4.5482E-04 3.9400E+06 9.2900E-04 3.9800E+06 6.0536E-04162533103 110 4.0400E+06 2.0134E-04 4.1000E+06 1.3697E-03 4.1900E+06 1.7611E-03162533103 111 4.3000E+06 6.2745E-04 4.4100E+06 5.0611E-04 4.4500E+06 4.0547E-04162533103 112 4.6000E+06 4.0395E-04 4.7500E+06 3.7820E-04 4.8200E+06 3.4763E-04162533103 113 4.8900E+06 3.0327E-04 5.0000E+06 2.5545E-04 5.2000E+06 3.9126E-04162533103 114 5.4000E+06 4.9800E-04 5.6000E+06 1.2711E-04 5.8000E+06 1.4110E-04162533103 115 6.0000E+06 7.1879E-05 6.5000E+06 2.8119E-05 7.0000E+06 8.8116E-06162533103 116 7.5000E+06 8.4801E-06 8.0000E+06 8.4522E-06 8.5000E+06 1.5461E-05162533103 117 9.0000E+06 1.7429E-05 9.5000E+06 2.0092E-05 1.0000E+07 1.8530E-05162533103 118 1.0500E+07 9.0846E-06 1.1000E+07 9.2760E-06 1.1500E+07 9.1510E-06162533103 119 1.2000E+07 1.6018E-05 1.2500E+07 1.3364E-05 1.3000E+07 1.1289E-05162533103 120 1.3500E+07 1.4055E-05 1.4000E+07 6.9831E-05 1.4500E+07 5.7844E-05162533103 121 1.5000E+07 4.2055E-05 1.6000E+07 3.8698E-05 1.7000E+07 2.3904E-05162533103 122 1.8000E+07 1.8309E-05 1.9000E+07 1.5467E-05 2.0000E+07 0.0000E+00162533103 123 0.0000E+00 0.0000E+00 0 1 118 59162533103 124 1.0000E-05 0.0000E+00 9.5700E+05 1.9119E-01 1.5800E+06 1.6485E-01162533103 125 1.6300E+06 5.1787E-02 1.6800E+06 8.5681E-02 1.7800E+06 1.9237E-01162533103 126 1.8300E+06 5.1277E-02 1.8800E+06 8.2068E-02 1.9200E+06 1.2620E-01162533103 127 1.9600E+06 1.7086E-01 2.0000E+06 2.4344E-01 2.0400E+06 3.6079E-01162533103 128 2.0800E+06 2.9571E-01 2.1200E+06 2.2469E-01 2.1600E+06 2.2179E-02162533103 129 2.2000E+06 8.2727E-03 2.2500E+06 1.5780E-02 2.3000E+06 1.3301E-02162533103 130 2.3500E+06 4.2311E-02 2.4000E+06 1.0882E-02 2.4500E+06 1.6003E-02162533103 131 2.5000E+06 2.6183E-02 2.5500E+06 1.0540E-02 2.6000E+06 2.6626E-02162533103 132 2.6500E+06 1.6542E-02 2.7000E+06 6.7118E-02 2.7500E+06 2.5660E-02162533103 133 2.8000E+06 1.0849E-02 2.8400E+06 1.5931E-02 2.9000E+06 1.3131E-02162533103 134 2.9500E+06 1.8754E-02 3.0000E+06 2.2004E-02 3.0500E+06 3.3742E-02162533103 135 3.1000E+06 2.6312E-02 3.1500E+06 3.8633E-02 3.2000E+06 2.2750E-02162533103 136 3.2400E+06 5.7436E-02 3.2800E+06 2.9198E-02 3.3300E+06 3.3861E-02162533103 137 3.3700E+06 2.2386E-02 3.4000E+06 1.9221E-02 3.4400E+06 2.5917E-02162533103 138 3.4700E+06 2.9828E-02 3.5100E+06 2.2332E-02 3.5500E+06 1.0696E-02162533103 139 3.6000E+06 2.2579E-02 3.6500E+06 1.6146E-02 3.7000E+06 3.0597E-02162533103 140 3.7500E+06 5.3481E-02 3.8000E+06 3.3928E-02 3.8500E+06 4.6404E-02162533103 141 3.9000E+06 8.2709E-02 3.9400E+06 1.4276E-01 3.9800E+06 6.9139E-02162533103 142 4.0400E+06 1.9062E-02 4.1000E+06 1.1884E-01 4.1900E+06 1.4459E-01162533103 143 4.3000E+06 5.3355E-02 5.2000E+06 0.0 162533103 144 0.0 0.0 1 5 210 20162533103 145 1.0000E-05 5.2000E+06 5.6000E+06 5.8000E+06 6.0000E+06 6.5000E+06162533103 146 7.0000E+06 8.5000E+06 9.0000E+06 9.5000E+06 1.0000E+07 1.0500E+07162533103 147 1.2000E+07 1.3500E+07 1.4000E+07 1.6000E+07 1.7000E+07 1.8000E+07162533103 148 1.9000E+07 2.0000E+07 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00162533103 149 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00162533103 150 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00162533103 151 0.0000E+00 0.0000E+00 0.0000E+00 2.6879E-01-1.6917E-04 7.6244E-05162533103 152 7.6222E-04 6.2428E-04 5.0197E-04 5.2466E-04 5.5936E-04 5.5109E-04162533103 153 5.3023E-04 5.2522E-04 4.1067E-04 3.3731E-04 7.0923E-04 7.5401E-04162533103 154 1.2360E-03 1.2362E-03 1.2365E-03 1.5941E-02 1.3651E-02 9.8297E-04162533103 155 6.7245E-04 5.9382E-04 6.0350E-04 6.3004E-04 6.2252E-04 6.0748E-04162533103 156 5.6306E-04 3.8830E-04 2.8840E-04 7.8421E-04 8.4368E-04 1.5155E-03162533103 157 1.5156E-03 1.5158E-03 1.6235E-02 9.6110E-04 6.8671E-04 5.8948E-04162533103 158 6.0468E-04 6.3316E-04 6.2636E-04 6.1117E-04 5.6685E-04 3.9017E-04162533103 159 2.9064E-04 7.9613E-04 8.5614E-04 1.5382E-03 1.5384E-03 1.5389E-03162533103 160 8.0479E-03 2.5416E-03 6.7123E-04 7.8051E-04 8.0128E-04 7.5961E-04162533103 161 7.2126E-04 6.5442E-04 4.3721E-04 3.8278E-04 5.5604E-04 5.8133E-04162533103 162 8.6206E-04 8.6210E-04 8.6223E-04 2.9649E-03 5.5056E-04 7.2580E-04162533103 163 7.6567E-04 7.1700E-04 6.7961E-04 6.3216E-04 4.3992E-04 3.8498E-04162533103 164 4.6967E-04 4.8689E-04 6.5527E-04 6.5528E-04 6.5537E-04 1.4913E-03162533103 165 6.4250E-04 6.5678E-04 6.2474E-04 5.9695E-04 5.2029E-04 3.8122E-04162533103 166 3.4843E-04 3.9503E-04 4.0583E-04 5.1030E-04 5.1045E-04 5.1050E-04162533103 167 1.4220E-03 7.3842E-04 6.8011E-04 7.0157E-04 6.0537E-04 4.1142E-04162533103 168 3.7075E-04 4.1240E-04 4.2421E-04 5.3492E-04 5.3508E-04 5.3511E-04162533103 169 1.5183E-03 9.1191E-04 6.9948E-04 6.0628E-04 4.2545E-04 3.8548E-04162533103 170 4.3714E-04 4.4949E-04 5.7461E-04 5.7469E-04 5.7485E-04 1.5774E-03162533103 171 7.0082E-04 6.9341E-04 4.1684E-04 3.6597E-04 4.2464E-04 4.3794E-04162533103 172 5.6898E-04 5.6912E-04 5.6922E-04 1.3354E-03 8.5551E-04 4.5574E-04162533103 173 3.6290E-04 4.0894E-04 4.2291E-04 5.4441E-04 5.4452E-04 5.4461E-04162533103 174 1.3514E-03 5.3505E-04 3.6596E-04 3.9729E-04 4.1346E-04 5.1956E-04162533103 175 5.1969E-04 5.1975E-04 1.6092E-03 3.8766E-04 3.3510E-04 3.4532E-04162533103 176 3.5854E-04 3.5862E-04 3.5868E-04 1.8445E-03-1.5721E-03-4.2629E-04162533103 177 2.6490E-04 2.6489E-04 2.6498E-04 1.3992E-02 5.9123E-03 7.7589E-04162533103 178 7.7619E-04 7.7627E-04 1.8117E-02 8.3496E-04 8.3499E-04 8.3513E-04162533103 179 1.8567E-02 1.3578E-02 1.3579E-02 2.0922E-02 1.3581E-02 2.3235E-02162533103 180 0.0000E+00 0.0000E+00 0 1 8 4162533103 181 1.0000E-05 0.0000E+00 9.2000E+05 1.9119E-01 9.5700E+05 0.0000E+00162533103 182 2.0000E+07 0.0000E+00 162533103 183 162533 099999 1625 0 0 0 0 0 0 0 2.10450E+4 4.45700E+1 2 0 1 12126 1451 1 0.0 0.0 0 0 0 62126 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022126 1451 3 3.00000E+2 0.0 1 0 104 42126 1451 4 21-Sc- 45 CNDC EVAL-FEB91 Z.X.ZHAO 2126 1451 5 DIST- 2126 1451 6 ----- DOSIMETRY FILE 2126 1451 7 ----- INCIDENT NEUTRON DATA 2126 1451 8 ----- ENDF-6 FORMAT 2126 1451 9 THIS EVALUATION IS MADE MAINLY BASED ON THE RESONANCE 2126 1451 10 PARAMETERS GIVEN BY REF.[1], MEASURED DATA OF REFS.[2-10] 2126 1451 11 AND THE SYSTEMATICS OF EXCITATION FUNCTION FOR (N,GAMMA) 2126 1451 12 REACTION OF REF.[11]. A SET OF RESOLVED RESONANCE PARAMETERS 2126 1451 13 ARE GIVEN FOR EN < 100 KEV. IN THE REGION OF EN > 100 KEV, 2126 1451 14 THE SMOOTH CROSS SECTIONS ARE GIVEN. 2126 1451 15 MF=2 MT=251 RESONANCE PARAMETER 2126 1451 16 POSITIVE ENERGY RESONANCE PARAMETERS LISTED BELOW ARE 2126 1451 17 BASED ON THOSE OF REF.[1]. TO FIT KENNY'S AVERAGE CROSS 2126 1451 18 SECTIONS [2] (CORRECTED IN REF.[3]), THE CAPTURE WIDTHS 2126 1451 19 HAVE BEEN ADJUSTED. THE NEGATIVE ENERGY LEVEL PARAMETERS 2126 1451 20 ARE TAKEN FROM JENDL-3 EVALUATION [12] TO REPRODUCE 2126 1451 21 THERMAL CROSS SECTION 27.2 B WHICH IS IN GOOD AGREEMENT 2126 1451 22 WITH 27.2 +- 0.2 B GIVEN BY REF.[1]. 2126 1451 23 2126 1451 24 MF=3 MT=102 SMOOTH CROSS SECTION 2126 1451 25 THE SMOOTH CROSS SECTION CALCULATED FROM ZHAO'S 2126 1451 26 SYSTEMATICS [11] ARE CONSISTENT WITH THE MEASURED DATA OF 2126 1451 27 REFS.[2-9] FROM 1 KEV TO 20 MEV. THEREFORE THE SYSTEMATICS 2126 1451 28 IS USED IN THIS EVALUATION TO GIVE THE SMOOTH CROSS 2126 1451 29 SECTIONS FOR EN > 100 KEV. 2126 1451 30 2126 1451 31 MF=8 MT=102 RADIOACTIVE DECAY DATA 2126 1451 32 HALF-TIME OF SC-46 IS TAKEN FROM REF.[13] 2126 1451 33 2126 1451 34 MF=32 AND 33 MT=102 COVARIANCE DATA 2126 1451 35 LONG-RANGE COMPONENT (GIVEN IN FILE 32): 2126 1451 36 10-5 EV TO 1 KEV----- THE ERROR OF THERMAL CROSS 2126 1451 37 SECTION. 2126 1451 38 1 KEV TO 100 KEV----- THE NORMALIZATION ERROR OF 2126 1451 39 KENNY'S MEASUREMENT. 2126 1451 40 100 KEV TO 20 MEV---- THE SHAPE ERROR OF THE 2126 1451 41 SYSTEMATICS. 2126 1451 42 SHORT-RANGE COMPONENT: 2126 1451 43 FOR EN < 100 KEV, THIS COMPONENT IS GIVEN IN FILE 32. 2126 1451 44 IN THIS FILE, ONLY THE ERRORS OF SOME LARGE S-WAVE 2126 1451 45 RESONANCE PARAMETERS ARE CONSIDERED. THE SHORT-RANGE 2126 1451 46 COMPONENT FOR EN > 100 KEV ARE GIVEN IN FILE 33 BASED ON 2126 1451 47 THE ERRORS OF PERKIN'S MEASUREMENT [6]. 2126 1451 48 2126 1451 49 REFERENCES 2126 1451 50 [1] S. F. MUGHABGHAB ET AL., BNL-325, 4TH, VOL.1 (1981) 2126 1451 51 [2] M. J. KENNY ET AL., AUST. J. PHYS., 30, 605 (1977) 2126 1451 52 [3] B. J. ALLEN ET AL., NUCL. SCI. AND ENG., 82, 230 (1982) 2126 1451 53 [4] S. A. ROMANOV ET AL., YF, 1, 229 (1965) 2126 1451 54 [5] R. BOOTH ET AL., PHYS. REV., 112, 226 (1958) 2126 1451 55 [6] J. L. PERKIN, J. NUCL. ENERGY, 17, 349 (1963) 2126 1451 56 [7] J. CSIKAI ET AL., NUCL. PHYS., A95, 229 (1967) 2126 1451 57 [8] M. WAGNER ET AL., APA, 52, 30 (1980) 2126 1451 58 [9] M. BUDNAR ET AL., INDC(YUG)-6 (1979) 2126 1451 59 [10] W. MANNHART, ZP/A, 272, 273 (1975) 2126 1451 60 [11] Z. X. ZHAO ET AL., CHINESE NUCL. PHYS.,VOL.11, 71 (1989)2126 1451 61 [12] T. NAKAGAWA ET AL., JAERI-M, 90-099 (1990) 2126 1451 62 [13] J. TULI, NUCLEAR WALLET CARDS, NNDC, 1990 2126 1451 63 2126 1451 64 ================================================================ 2126 1451 65 The Q values and threshold energies were updated prior to pro- 2126 1451 66 cessing through the codes to comply with the values obtained 2126 1451 67 using the NNDC calculation program which is based on the 1995 2126 1451 68 Update to the Atomic mass Evaluation. 2126 1451 69 2126 1451 70 Total and Elastic cross sections were only given upto 100 Kev 2126 1451 71 in the original data.For these two reactions the energy region 2126 1451 72 between 100 KeV and 20 Mev was taken the ENDF/B-6 library, 2126 1451 73 linearised and added to this file. 2126 1451 74 2126 1451 75 The original SC45G cross section set had uncertainties for the 2126 1451 76 resonance parameters in file MF=32, which were converted in the 2126 1451 77 extended SAND-II group structure and inserted in file MF=33 as 2126 1451 78 extra "NI type" sub-subsections. 2126 1451 79 2126 1451 80 ***************** Program LINEAR (VERSION 2002-1) ***************2126 1451 81 For All Data Greater than 1.0000E-10 barns in Absolute Value 2126 1451 82 Data Linearized to Within an Accuracy of .100000000 per-cent 2126 1451 83 ***************** Program RECENT (VERSION 2002-1) ***************2126 1451 84 for All Data Greater than 1.0000E-10 barns in Absolute Value 2126 1451 85 Data Linearized to within an Accuracy of .100000000 per-cent 2126 1451 86 ***************** Program SIGMA1 (VERSION 2002-1) ***************2126 1451 87 Data Doppler Broadened to 300.000000 Kelvin 2126 1451 88 for All Data Greater than 1.0000E-10 barns in Absolute Value 2126 1451 89 Data Linearized to Within an Accuracy pf .100000000 per-cent 2126 1451 90 ***************** Program FIXUP (Version 2002-1) ****************2126 1451 91 Corrected ZA/AWR in All Sections-----------------------------Yes 2126 1451 92 Corrected Thresholds-----------------------------------------Yes 2126 1451 93 Extended Cross Sections to 20 MeV----------------------------No 2126 1451 94 Allow Cross Section Deletion---------------------------------No 2126 1451 95 Allow Cross Section Reconstruction---------------------------No 2126 1451 96 Make All Cross Sections Non-Negative-------------------------Yes 2126 1451 97 Delete Energies Not in Ascending Order-----------------------Yes 2126 1451 98 Deleted Duplicate Points-------------------------------------Yes 2126 1451 99 Check for Ascending MAT/MF/MT Order--------------------------Yes 2126 1451 100 Check for Legal MF/MT Numbers--------------------------------Yes 2126 1451 101 Allow Creation of Missing Sections---------------------------No 2126 1451 102 Allow Insertion of Energy Points-----------------------------No 2126 1451 103 Create Uniform Energy Grid-----------------------------------No 2126 1451 104 Delete Section if Cross Section =0 at All Energies-----------Yes 2126 1451 105 ***************** Program GROUPIE (VERSION 2002-1) **************2126 1451 106 Unshielded Group Averages Using 640 Groups 2126 1451 107 Weighting Spectrum: Flat (Constant) Spectrum 2126 1451 108 1 451 112 12126 1451 109 2 151 201 12126 1451 110 3 102 217 12126 1451 111 33 102 150 12126 1451 112 2126 1 099999 2126 0 0 0 2.10450E+4 4.45700E+1 0 0 1 02126 2151 1 2.10450E+4 1.00000E+0 0 0 1 02126 2151 2 1.00000E-5 1.00000E+5 1 2 0 02126 2151 3 3.50000E+0 4.55000E-1 0 0 2 02126 2151 4 4.45700E+1 0.0 0 0 492 822126 2151 5 -700.000000 3.00000000 107.110000 106.700000 .410000000 0.0 2126 2151 6 -294.000000 4.00000000 30.6660000 30.2600000 .405500000 0.0 2126 2151 7 3295.00000 3.00000000 75.7500000 75.0000000 .750350000 0.0 2126 2151 8 4330.00000 4.00000000 340.770000 340.000000 .768540000 0.0 2126 2151 9 6692.00000 3.00000000 130.900000 130.000000 .904560000 0.0 2126 2151 10 8023.00000 4.00000000 145.640000 145.000000 .640450000 0.0 2126 2151 11 9080.00000 3.00000000 291.060000 290.000000 1.05870000 0.0 2126 2151 12 10662.0000 3.00000000 14.8180000 14.0000000 .818250000 0.0 2126 2151 13 10740.0000 4.00000000 15.7910000 15.0000000 .790940000 0.0 2126 2151 14 11580.0000 4.00000000 290.770000 290.000000 .772110000 0.0 2126 2151 15 14525.0000 3.00000000 20.5720000 20.0000000 .572490000 0.0 2126 2151 16 14820.0000 4.00000000 26.4670000 26.0000000 .467030000 0.0 2126 2151 17 15560.0000 4.00000000 28.6020000 28.0000000 .601680000 0.0 2126 2151 18 15850.0000 3.00000000 5.79090000 5.00000000 .790940000 0.0 2126 2151 19 18580.0000 3.00000000 32.7910000 32.0000000 .790940000 0.0 2126 2151 20 18870.0000 4.00000000 62.6980000 62.0000000 .697730000 0.0 2126 2151 21 20500.0000 4.00000000 80.9360000 80.0000000 .935510000 0.0 2126 2151 22 20930.0000 3.00000000 750.670000 750.000000 .674650000 0.0 2126 2151 23 24010.0000 3.00000000 51.2260000 50.0000000 1.22570000 0.0 2126 2151 24 24315.0000 4.00000000 70.5200000 70.0000000 .519610000 0.0 2126 2151 25 26925.0000 4.00000000 103.010000 102.000000 1.01200000 0.0 2126 2151 26 27900.0000 3.00000000 110.910000 110.000000 .912460000 0.0 2126 2151 27 29480.0000 4.00000000 30.6910000 30.0000000 .691420000 0.0 2126 2151 28 29630.0000 4.00000000 50.7230000 50.0000000 .722840000 0.0 2126 2151 29 32180.0000 3.00000000 571.150000 570.000000 1.15150000 0.0 2126 2151 30 33730.0000 3.00000000 201.090000 200.000000 1.09270000 0.0 2126 2151 31 34860.0000 3.00000000 114.990000 114.000000 .991700000 0.0 2126 2151 32 35080.0000 4.00000000 150.690000 150.000000 .685020000 0.0 2126 2151 33 39980.0000 4.00000000 181.350000 180.000000 1.35010000 0.0 2126 2151 34 40500.0000 3.00000000 89.3740000 88.0000000 1.37400000 0.0 2126 2151 35 40815.0000 4.00000000 124.690000 124.000000 .687000000 0.0 2126 2151 36 43050.0000 4.00000000 151.150000 150.000000 1.14700000 0.0 2126 2151 37 43215.0000 3.00000000 31.1530000 30.0000000 1.15300000 0.0 2126 2151 38 45730.0000 3.00000000 420.560000 420.000000 .560000000 0.0 2126 2151 39 47180.0000 3.00000000 60.5700000 60.0000000 .570000000 0.0 2126 2151 40 48790.0000 4.00000000 40.5130000 40.0000000 .513000000 0.0 2126 2151 41 49120.0000 3.00000000 40.4420000 40.0000000 .442000000 0.0 2126 2151 42 51160.0000 4.00000000 841.400000 840.000000 1.39600000 0.0 2126 2151 43 51685.0000 3.00000000 40.4320000 40.0000000 .432000000 0.0 2126 2151 44 52025.0000 4.00000000 40.4970000 40.0000000 .497000000 0.0 2126 2151 45 52330.0000 3.00000000 70.4490000 70.0000000 .449000000 0.0 2126 2151 46 53075.0000 3.00000000 40.3040000 40.0000000 .304000000 0.0 2126 2151 47 54730.0000 3.00000000 41.8900000 40.0000000 1.89000000 0.0 2126 2151 48 55125.0000 3.00000000 40.4780000 40.0000000 .478000000 0.0 2126 2151 49 57890.0000 3.00000000 35.5230000 35.0000000 .523000000 0.0 2126 2151 50 58770.0000 3.00000000 1435.80000 1435.00000 .840000000 0.0 2126 2151 51 59110.0000 3.00000000 51.1240000 50.0000000 1.12400000 0.0 2126 2151 52 60140.0000 3.00000000 40.8350000 40.0000000 .835000000 0.0 2126 2151 53 61890.0000 4.00000000 520.750000 520.000000 .751000000 0.0 2126 2151 54 62490.0000 3.00000000 495.070000 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85 2.9936E-08 1.3649E-08 9.2255E-08 8.4481E-08 1.1418E-08 8.4861E-08212633102 86 3.6933E-08 5.7820E-06 3.3369E-05 3.0616E-04 5.7305E-06 8.0793E-06212633102 87 9.4646E-06 2.7073E-05 4.5174E-05 1.6515E-05 5.2597E-06 2.4645E-06212633102 88 1.3982E-06 8.7856E-07 2.2830E-07 4.2504E-07 6.6183E-07 1.9753E-06212633102 89 3.0563E-07 1.4544E-07 8.5161E-07 7.5803E-07 1.0883E-07 7.4987E-07212633102 90 3.3273E-07 2.0963E-04 1.9379E-03 3.4155E-05 3.4355E-05 2.3371E-05212633102 91 4.3038E-05 6.6431E-05 2.5386E-05 8.9490E-06 4.5915E-06 2.7974E-06212633102 92 1.8542E-06 4.9895E-07 9.2670E-07 1.0662E-06 2.4945E-06 5.8495E-07212633102 93 3.1928E-07 1.1255E-06 9.7470E-07 1.7988E-07 8.8768E-07 4.3584E-07212633102 94 1.8310E-02 3.2576E-04 2.8938E-04 1.3803E-04 1.1212E-04 1.2320E-04212633102 95 5.7938E-05 2.8618E-05 1.8113E-05 1.2562E-05 9.0537E-06 2.5688E-06212633102 96 4.8481E-06 3.9722E-06 5.1392E-06 2.7835E-06 1.7502E-06 2.8066E-06212633102 97 2.5018E-06 7.5487E-07 1.7106E-06 1.1796E-06 6.0844E-06 6.9310E-06212633102 98 6.2869E-06 1.5456E-05 2.5212E-05 9.3414E-06 3.0736E-06 1.4842E-06212633102 99 8.6228E-07 5.5061E-07 1.4373E-07 2.5891E-07 2.7967E-07 6.2654E-07212633102 100 1.5322E-07 8.3156E-08 2.7751E-07 2.2733E-07 4.3656E-08 2.0351E-07212633102 101 1.0156E-07 2.1420E-05 3.9156E-05 1.3226E-04 2.2500E-04 8.1404E-05212633102 102 2.5305E-05 1.1547E-05 6.3848E-06 3.9080E-06 9.8334E-07 1.7063E-06212633102 103 1.7295E-06 3.7268E-06 9.3236E-07 4.9759E-07 1.6066E-06 1.2634E-06212633102 104 2.4505E-07 1.1182E-06 5.5889E-07 8.2376E-05 2.8835E-04 4.9153E-04212633102 105 1.7730E-04 5.4890E-05 2.4924E-05 1.3711E-05 8.3456E-06 2.0843E-06212633102 106 3.5510E-06 3.0519E-06 5.2665E-06 1.7699E-06 9.9046E-07 2.3172E-06212633102 107 1.8166E-06 4.2118E-07 1.4473E-06 8.0185E-07 1.0186E-03 1.7344E-03212633102 108 6.2471E-04 1.9346E-04 8.7819E-05 4.8271E-05 2.9334E-05 7.3006E-06212633102 109 1.2276E-05 8.9112E-06 1.0675E-05 5.6377E-06 3.3209E-06 5.0173E-06212633102 110 4.1306E-06 1.2302E-06 2.6350E-06 1.8117E-06 2.9922E-03 1.0875E-03212633102 111 3.3573E-04 1.5214E-04 8.3555E-05 5.0747E-05 1.2622E-05 2.1180E-05212633102 112 1.4928E-05 1.6361E-05 9.5985E-06 5.7041E-06 7.8220E-06 6.5105E-06212633102 113 2.0601E-06 3.8693E-06 2.8536E-06 3.9777E-04 1.2262E-04 5.5543E-05212633102 114 3.0517E-05 1.8559E-05 4.6303E-06 7.8604E-06 6.4671E-06 1.0325E-05212633102 115 3.8327E-06 2.1689E-06 4.3923E-06 3.1927E-06 8.5061E-07 2.3383E-06212633102 116 1.4256E-06 3.7889E-05 1.7207E-05 9.4876E-06 5.8048E-06 1.4671E-06212633102 117 2.6099E-06 3.3811E-06 9.0932E-06 1.6277E-06 7.8228E-07 3.3150E-06212633102 118 1.9151E-06 3.7767E-07 1.8012E-06 8.8737E-07 7.8447E-06 4.3534E-06212633102 119 2.6974E-06 7.0139E-07 1.3760E-06 3.0715E-06 1.0712E-05 1.2243E-06212633102 120 4.6912E-07 3.4699E-06 1.6492E-06 2.7246E-07 1.7667E-06 7.9571E-07212633102 121 2.4461E-06 1.5543E-06 4.2703E-07 9.8619E-07 3.5720E-06 1.3965E-05212633102 122 1.2612E-06 3.8683E-07 3.9869E-06 1.5685E-06 2.4285E-07 1.8345E-06212633102 123 7.9547E-07 1.0380E-06 3.1496E-07 9.1294E-07 4.7837E-06 1.9682E-05212633102 124 1.5729E-06 3.9925E-07 4.7579E-06 1.5139E-06 2.3795E-07 1.9312E-06212633102 125 8.2324E-07 1.1255E-07 4.2344E-07 2.8478E-06 1.1982E-05 8.9628E-07212633102 126 1.9429E-07 2.2608E-06 5.5343E-07 9.5443E-08 7.8940E-07 3.3449E-07212633102 127 2.0742E-06 1.6466E-05 6.9671E-05 4.9909E-06 9.2554E-07 7.9133E-06212633102 128 1.3090E-06 2.8949E-07 2.2393E-06 9.5624E-07 1.4433E-04 5.9024E-04212633102 129 4.0501E-05 6.5832E-06 1.5000E-05 2.4583E-06 7.9553E-07 3.1192E-06212633102 130 1.5155E-06 2.5686E-03 1.8228E-04 2.8441E-05 3.1009E-05 8.8389E-06212633102 131 2.6695E-06 5.5532E-06 3.2812E-06 1.3371E-05 2.5234E-06 4.0262E-05212633102 132 6.1624E-06 5.0010E-07 4.1452E-06 1.7801E-06 1.8179E-06 1.4569E-04212633102 133 5.2352E-05 5.4077E-08 4.7205E-06 1.9613E-06 1.7427E-02 9.9573E-03212633102 134 8.4587E-05 7.9004E-06 3.3483E-06 9.0737E-03 8.8884E-05 4.9176E-05212633102 135 2.2947E-05 4.7205E-06 4.3502E-05 1.7438E-05 4.8803E-04 1.9935E-04212633102 136 8.4828E-05 212633102 137 0.0 0.0 1 5 10 4212633102 138 1.1000E+04 1.1500E+04 1.2000E+04 1.2750E+04 3.0436E-05 1.0200E-04212633102 139 7.2626E-06 3.8928E-04 2.8587E-05 2.1192E-06 212633102 140 0.0 0.0 1 5 15 5212633102 141 2.3000E+04 2.4000E+04 2.5500E+04 2.7000E+04 2.8000E+04 2.0221E-03212633102 142 1.1416E-04 2.6389E-07 8.7044E-08 1.0575E-03 8.1002E-07 1.7944E-07212633102 143 6.8298E-04 3.0746E-05 9.1324E-05 212633102 144 0.0 0.0 0 1 12 6212633102 145 3.4000E+04 1.5819E-05 3.6000E+04 4.9680E-12 3.8000E+04 6.5496E-05212633102 146 4.0000E+04 4.8828E-05 4.2500E+04 1.6762E-05 4.5000E+04 0.0000E+00212633102 147 0.0 0.0 0 1 12 6212633102 148 6.9000E+04 1.6282E-05 7.2000E+04 2.4062E-09 7.6000E+04 2.6452E-08212633102 149 8.0000E+04 1.7513E-08 8.4000E+04 2.4018E-04 8.8000E+04 0.0000E+00212633102 150 212633 099999 2126 0 0 0 0 0 0 0 2.20460E+4 4.55579E+1 0 0 34 102225 1451 1 0.0 0.0 0 0 0 62225 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022225 1451 3 3.00000E+2 0.0 1 0 233 62225 1451 4 22-Ti- 46 FEI EVAL-Jan02 K.I.Zolotarev 2225 1451 5 DIST-Feb2004 2225 1451 6 ----IRDF-2002 MATERIAL 2225 2225 1451 7 -----INCIDENT NEUTRON DATA 2225 1451 8 ------ENDF-6 FORMAT 2225 1451 9 ***************************************************************** 2225 1451 10 22-TI- 46 FEI EVAL-Jan02 K.I.Zolotarev 2225 1451 11 DIST-Feb02 20020203 2225 1451 12 ----BROND-2 MATERIAL 2225 2225 1451 13 -----INCIDENT NEUTRON DATA 2225 1451 14 ------ENDF-6 FORMAT 2225 1451 15 ------Russian Reactor Dosimetry File RRDF-2002 2225 1451 16 ***************************************************************** 2225 1451 17 Author of evaluation: K.I.Zolotarev 2225 1451 18 ***************************************************************** 2225 1451 19 MF=3 2225 1451 20 MT= 16 -(n,2n) cross section 2225 1451 21 ------------------------------------- 2225 1451 22 Excitation function for the Ti-46(n,2n)Ti-45 reaction in the 2225 1451 23 energy region from threshold to 20 MeV was evaluated by means of 2225 1451 24 statistical analysis of experimental cross section data [1-18]. 2225 1451 25 All experimental data were renormalized to the new standards 2225 1451 26 for monitor reactions cross sections and decay data. In the measu-2225 1451 27 rements [5,14-16] uncertainties were increased until the data had 2225 1451 28 become mutually consistent. 2225 1451 29 The final procedure of evaluation the Ti-46(n,2n)Ti-45 reac- 2225 1451 30 tion excitation function has been carried out within the frame- 2225 1451 31 work of generalized least squares method. Statistical analysis of 2225 1451 32 input cross section data was carried out by means of PADE-2 code 2225 1451 33 [19]. Rational function was used as the model function [20]. 2225 1451 34 The evaluated Ti-46(n,2n)Ti-45 excitation function averaged 2225 1451 35 on U-235 thermal neutron fission spectrum [21] and Cf-252 sponta- 2225 1451 36 neous fission neutron spectrum [22] gives the following values : 2225 1451 37 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 38 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2225 1451 39 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 40 U-235 neutron fission ³ 0.0044686 ³ 2225 1451 41 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 42 CF-252 spont. fission ³ 0.011982 ³ 0.093 +- 0.031 [23] 2225 1451 43 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 44 2225 1451 45 MT=103 -(n,p) cross section 2225 1451 46 ------------------------------------- 2225 1451 47 Excitation function for the Ti-46(n,p)Sc-46m+g reaction in 2225 1451 48 the energy region from threshold to 20 MeV was evaluated by means 2225 1451 49 of statistical analysis of experimental cross section data [6,17, 2225 1451 50 24-38] and data from GNASH calculation [39]. 2225 1451 51 Analyzed microscopic experimental data were renormalized to 2225 1451 52 the new recommended standards for monitor reaction cross sections 2225 1451 53 and decay data. All experiments performed with Ti-natural samples 2225 1451 54 were corrected for the contribution of the Ti-47(n,np+pn)Sc-46m+g 2225 1451 55 and Ti-47(n,d)Sc-46m+g reactions. Total cross section for these 2225 1451 56 reactions were taken from ref.[40]. Experimental data from ref. 2225 1451 57 [6,27] were used partially. Data of Bormann et al. [6] were taken 2225 1451 58 only for 14.1 MeV point. In the work of Lukic and Carroll [27] it 2225 1451 59 were used data measured relative Fe56(n,p)Mn56 and Al27(n,a)Na24 2225 1451 60 monitor reactions. These data were renormalized to the cross sec- 2225 1451 61 sections obtained in the ref.[28]. Data of Smith and Meadows [28] 2225 1451 62 measured with using D(d,n)He3 neutron source were renormalized 2225 1451 63 in the energy range 6.996-9.950 MeV to the preliminary evaluated 2225 1451 64 integral of cross section from new experimental data [34,37,38]. 2225 1451 65 Experimental cross section data [8] , [41-46] were rejected 2225 1451 66 due to their discrepancy with the main bulk of experimental data 2225 1451 67 [6,17,24-38] and data from theoretical model calculation. 2225 1451 68 Statistical analysis of input cross section data was carried 2225 1451 69 out by means of PADE-2 code [19]. Rational function was used as 2225 1451 70 the model function [20]. 2225 1451 71 U-235 thermal fission [21] and Cf-252 spontaneous fission 2225 1451 72 neutron spectra [22] averaged cross sections calculated from the 2225 1451 73 evaluated Ti-46(n,p)Sc-46m+g reaction excitation function are 2225 1451 74 the following: 2225 1451 75 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 76 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2225 1451 77 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 78 U-235 neutron fission ³ 11.447 ³ 11.57 +- 0.37 [47] 2225 1451 79 ³ ³ 11.51 +- 0.40 [48] 2225 1451 80 ³ ³ 11.55 +- 0.20 [49] 2225 1451 81 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 82 CF-252 spont. fission ³ 13.818 ³ 14.20 +- 0.24 [50] 2225 1451 83 ³ ³ 14.07 +- 0.25 [51] 2225 1451 84 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2225 1451 85 2225 1451 86 MF=33 2225 1451 87 MT= 16 -(n,2n) cross section cov. matrix 2225 1451 88 ---------------------------------------- 2225 1451 89 Uncertainties in the evaluated excitation function for the 2225 1451 90 reaction Ti-46(n,2n)Ti-45 are given in the form of relative cova- 2225 1451 91 riance matrix for the 17-neutron energy groups (LB=5). Covariance 2225 1451 92 matrix of uncertainties was calculated simultaneously with 2225 1451 93 recommended cross section data by means of PADE-2 code. 2225 1451 94 Eigenvalues of the 6-th digits relative covariance matrix 2225 1451 95 given in the 33-file are the following: 2225 1451 96 2225 1451 97 2.26418E-07 2.55017E-07 2.97541E-07 3.62667E-07 2225 1451 98 4.74369E-07 6.90723E-07 1.00928E-06 1.38635E-06 2225 1451 99 2.05902E-06 3.42154E-06 6.73918E-06 1.77794E-05 2225 1451 100 1.16721E-04 5.82882E-03 7.49965E-03 2.28734E-02 2225 1451 101 4.20028E-02 2225 1451 102 2225 1451 103 MT=103 -(n,p) cross section cov. matrix 2225 1451 104 ---------------------------------------- 2225 1451 105 Uncertainties in the evaluated excitation function for the 2225 1451 106 reaction Ti-46(n,p)Sc-45m+g are given in the form of relative co- 2225 1451 107 variance matrix for the 31-neutron energy groups (LB=5). Covari- 2225 1451 108 ance matrix of uncertainties was calculated simultaneously with 2225 1451 109 recommended cross section data by means of PADE-2 code. 2225 1451 110 Eigenvalues of the 6-th digits relative covariance matrix 2225 1451 111 given in the 33-file are the following: 2225 1451 112 2225 1451 113 3.10934E-07 3.16123E-07 3.24666E-07 3.37022E-07 2225 1451 114 3.51765E-07 3.72727E-07 3.91781E-07 4.23365E-07 2225 1451 115 4.47112E-07 4.91264E-07 5.29779E-07 6.15458E-07 2225 1451 116 6.75168E-07 7.93660E-07 9.79888E-07 1.29384E-06 2225 1451 117 1.98373E-06 3.65835E-06 2.36258E-04 5.76816E-04 2225 1451 118 7.66259E-04 1.00346E-03 1.10942E-03 1.43365E-03 2225 1451 119 2.89439E-03 3.32523E-03 4.20534E-03 7.47772E-03 2225 1451 120 1.06722E-02 1.67016E-02 7.87709E-02 2225 1451 121 2225 1451 122 References : 2225 1451 123 1. A.Poularikas, R.W.Fink Physical Review, v.115, p.989, 1959 2225 1451 124 2. R.J.Prestwood,B.P.Bayhurst Phys. Rev, v.121, p.1438,Mar.1961 2225 1451 125 3. L.A.Rayburn Phys. Rev., v.122, p.168, 1961 2225 1451 126 3. M.Cevolani, S.Petralia Nuovo Cimento, v.26, p.1328, Dec. 1962 2225 1451 127 5. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 2225 1451 128 6. M.Bormann et al. Nuclear Physics, v.63, p.438, March 1965 2225 1451 129 7. J.Csikai Progress Report EANDC-50S, v.2, p.102, July 1965 2225 1451 130 8. H.L.Pai Canadian J. of Physics, v.44, p.2337, 1966 2225 1451 131 9. G.N.Maslov et al. Yadernye Konstanty, No.9, p.50, 1972 2225 1451 132 10. J.Araminowicz, J.Dresler Report INR-1464, p.14, May 1973 2225 1451 133 11. A.Paulsen et al. Atomkernenergie , v.26, p.34, August 1975 2225 1451 134 12. S.M.Qaim, N.I.Molla Proc of 9th Symposium on Fusion Technolo-2225 1451 135 gy, Garmisch, Germany, 14-18 June 1976, p.589 2225 1451 136 13. R.A.Sigg J. Diss. Abst., v.B37, p.2237, November 1976 2225 1451 137 14. J.Csikai Proc. of Int. Conf. on Nuclear Data for Science and 2225 1451 138 Technology, Antwerp, 6-10 September 1982, p.414, Reidel Publ. 2225 1451 139 Company, 1983 2225 1451 140 15. N.T.Molla et al. Progress Report INDC(BAN)-002, p.1, Feb.1983 2225 1451 141 16. Zhou Muyao et al. Chinese J. of Nucl. Phys., v.9, p.34, 1987 2225 1451 142 17. Y.Ikeda et al. Report JAERI-1312, March 1988 2225 1451 143 18. P.M.Dighe e.a. Indian J. of Pure and Applied Physics, v.29, 2225 1451 144 p.665, October 1991 2225 1451 145 19. S.A.Badikov et.al. Preprint FEI-1686, Obninsk, 1985 2225 1451 146 20. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2225 1451 147 st's Meeting on Evaluation and Processing of Covariance Data, 2225 1451 148 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2225 1451 149 21. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2225 1451 150 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2225 1451 151 22. W.Mannhart IAEA-TECDOC-410, p.158, 1987 2225 1451 152 23. J.Csikai, Z.Dezso Proc. of 4th All Union Conf. on Neutron 2225 1451 153 Physics, Kiev, 18-22 April 1977, v.3, p.32 2225 1451 154 24. H.Liskien, A.Paulsen Nucl. Phys., v.63, p.393, March 1965 2225 1451 155 25. V.N.Levkovskij et al. Yadernaja Fizika (Sov.), v.10, n.1, 2225 1451 156 p.44, July 1969 2225 1451 157 26. S.K.Ghorai et al. J. Nucl. Energ., v.25, p.319, August 1971 2225 1451 158 27. Y.Lukic, E.E.Carroll Nucl. Sci. Eng., v.43, p.233, 1971 2225 1451 159 28. D.L.Smith, J.W.Meadows Nucl. Sci. Eng., v.58, p.314, 1975 2225 1451 160 29. Lu Hamlin et al. Chinese J. of Atomic Energy Sci. Technology,2225 1451 161 v.9, no.2, p.113, May 1975 2225 1451 162 30. K.Kayashima et al. Prog. Report NEANDC(J)-61U, p.94, Sep.1979 2225 1451 163 31. M.Viennot et al. Proc of Int. Conference on Nuclear Data for 2225 1451 164 Science and Technology, Antwerp, 6-10 September 1982, p.406 2225 1451 165 32. N.I.Molla et al. Progress Report INDC(BAN)-003, Sept. 1986 2225 1451 166 33. Lu Hanlin et al. Report INDC(CPR)-16, IAEA, August 1989 2225 1451 167 34. N.I.Molla, S.M.Qaim, M.Uhl Phys. Rev. C, v.42, n.4, p.1540, 2225 1451 168 October 1990 2225 1451 169 35. M.Viennot et al. Nucl. Sci. Eng., v.108, p.289, July 1991 2225 1451 170 36. Yuan Junqian e.a. High Energy Physics and Nucl.Phys.(China), 2225 1451 171 v.16, n.1, p.57, January 1992 2225 1451 172 37. J.W.Meadows, D.L.Smith, L.R.Greenwood, R.C.Haight, Y.Ikeda, 2225 1451 173 C.Konno Annals of Nuclear Energy, v.23, p.877, July 1996 2225 1451 174 38. Lu Hanlin et al. Report INDC(CPR)-045, IAEA, October 1998 2225 1451 175 39. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 2225 1451 176 Model Code for Calculation of Cross Section and Emission 2225 1451 177 Spectra. Report LA-6947, Los Alamos, 1977 ; 2225 1451 178 E.L.Trykiv, G.Ya.Tertychnyi Private communication, IPPE, 2225 1451 179 Obninsk, May 1999 2225 1451 180 40. K.I.Zolotrarev RRDF-2002, MAT=2222, eval. January 2002 2225 1451 181 41. D.L.Allan Nucl. Phys., v.24, p.274, April 1961 2225 1451 182 42. W.G.Cross, H.L.Pai Prog. Report EANDC(CAN)-16, p.1, Jan. 1963 2225 1451 183 43. S.M.Qaim, N.I.Molla Nucl. Phys., v.A283, p.269, June 1977 2225 1451 184 44. I.Ribansky, S.Gmuca J. Phys. G, v.9, p.1537, December 1983 2225 1451 185 45. R.Pepelnik et al. Progress Report, NEANDC(E)-262U,(5), p.32, 2225 1451 186 June 1985 2225 1451 187 46. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 2225 1451 188 47. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2225 1451 189 Washington D.C., 25-28 April 1989, v.2, p.923 2225 1451 190 48. W.Mannhart Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry, 2225 1451 191 Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985 2225 1451 192 49. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 2225 1451 193 50. W.Mannhart Handbook on Nuclear Activation Data . IAEA Tech- 2225 1451 194 nical Report series No.273, p.413, 1987 2225 1451 195 51. W.Mannhart Validation of Differential Cross Sections with 2225 1451 196 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2225 1451 197 September 2002 2225 1451 198 ***************************************************************** 2225 1451 199 The Q values and threshold energies were updated prior to pro- 2225 1451 200 cessing through the codes to comply with the values obtained 2225 1451 201 using the NNDC calculation program which is based on the 1995 2225 1451 202 Update to the Atomic mass Evaluation. 2225 1451 203 2225 1451 204 File 2 added to the pointwise file containing only the effective 2225 1451 205 scattering radius with no resonance parameters given. 2225 1451 206 Taken from ENDF/B-VI 2225 1451 207 2225 1451 208 Psuedo Threshold of 0.0 added at 2.1E+6 ev to MF/MT=3/103 2225 1451 209 in original evaluation 2225 1451 210 2225 1451 211 ***************************************************************** 2225 1451 212 ***************** Program LINEAR (VERSION 2002-1) ***************2225 1451 213 For All Data Greater than 1.0000E-10 barns in Absolute Value 2225 1451 214 Data Linearized to Within an Accuracy of .100000000 per-cent 2225 1451 215 ***************** Program SIGMA1 (VERSION 2002-1) ***************2225 1451 216 Data Doppler Broadened to 300.000000 Kelvin 2225 1451 217 for All Data Greater than 1.0000E-10 barns in Absolute Value 2225 1451 218 Data Linearized to Within an Accuracy pf .100000000 per-cent 2225 1451 219 ***************** Program FIXUP (Version 2002-1) ****************2225 1451 220 Corrected ZA/AWR in All Sections-----------------------------Yes 2225 1451 221 Corrected Thresholds-----------------------------------------Yes 2225 1451 222 Extended Cross Sections to 20 MeV----------------------------No 2225 1451 223 Allow Cross Section Deletion---------------------------------No 2225 1451 224 Allow Cross Section Reconstruction---------------------------No 2225 1451 225 Make All Cross Sections Non-Negative-------------------------Yes 2225 1451 226 Delete Energies Not in Ascending Order-----------------------Yes 2225 1451 227 Deleted Duplicate Points-------------------------------------Yes 2225 1451 228 Check for Ascending MAT/MF/MT Order--------------------------Yes 2225 1451 229 Check for Legal MF/MT Numbers--------------------------------Yes 2225 1451 230 Allow Creation of Missing Sections---------------------------No 2225 1451 231 Allow Insertion of Energy Points-----------------------------No 2225 1451 232 Create Uniform Energy Grid-----------------------------------No 2225 1451 233 Delete Section if Cross Section =0 at All Energies-----------Yes 2225 1451 234 ***************** Program GROUPIE (VERSION 2002-1) **************2225 1451 235 Unshielded Group Averages Using 640 Groups 2225 1451 236 Weighting Spectrum: Flat (Constant) Spectrum 2225 1451 237 1 451 243 12225 1451 238 2 151 4 12225 1451 239 3 16 26 12225 1451 240 3 103 63 12225 1451 241 33 16 35 12225 1451 242 33 103 97 12225 1451 243 2225 1 099999 2225 0 0 0 2.20460E+4 4.55578E+1 0 0 1 02225 2151 1 2.204600+4 1.000000+0 0 0 1 02225 2151 2 1.000000-5 1.000000+5 0 0 0 02225 2151 3 0.000000+0 0.573300+0 0 0 0 02225 2151 4 2225 2 099999 2225 0 0 0 2.20460E+4 4.55579E+1 0 0 0 02225 3 16 1 -1.31898E+7-1.31898E+7 0 0 1 672225 3 16 2 67 1 2225 3 16 3 13400000.0 5.44722E-6 13500000.0 .000269054 13600000.0 .0013538942225 3 16 4 13700000.0 .003167185 13800000.0 .005539635 13900000.0 .0084628702225 3 16 5 14000000.0 .011885800 14100000.0 .015756100 14200000.0 .0200212002225 3 16 6 14300000.0 .024629400 14400000.0 .029530600 14500000.0 .0346768502225 3 16 7 14600000.0 .040022950 14700000.0 .045526750 14800000.0 .0511660252225 3 16 8 14900000.0 .056872075 15000000.0 .062608313 15100000.0 .0683747382225 3 16 9 15200000.0 .074141163 15300000.0 .079907588 15400000.0 .0855952332225 3 16 10 15500000.0 .091204100 15600000.0 .096812967 15700000.0 .1022853332225 3 16 11 15800000.0 .107621200 15900000.0 .112957067 16000000.0 .1181235002225 3 16 12 16100000.0 .123120500 16200000.0 .128117500 16300000.0 .1329310002225 3 16 13 16400000.0 .137561000 16500000.0 .142191000 16600000.0 .1466356672225 3 16 14 16700000.0 .150895000 16800000.0 .155154333 16900000.0 .1592341672225 3 16 15 17000000.0 .163134500 17100000.0 .167034833 17200000.0 .1707660002225 3 16 16 17300000.0 .174328000 17400000.0 .177890000 17500000.0 .1812951672225 3 16 17 17600000.0 .184543500 17700000.0 .187791833 17800000.0 .1908743752225 3 16 18 17900000.0 .193791125 18000000.0 .196707875 18100000.0 .1996246252225 3 16 19 18200000.0 .202373250 18300000.0 .204953750 18400000.0 .2075342502225 3 16 20 18500000.0 .210114750 18600000.0 .212532000 18700000.0 .2147860002225 3 16 21 18800000.0 .217040000 18900000.0 .219294000 19000000.0 .2215480002225 3 16 22 19100000.0 .223648400 19200000.0 .225595200 19300000.0 .2275420002225 3 16 23 19400000.0 .229488800 19500000.0 .231435600 19600000.0 .2332651252225 3 16 24 19700000.0 .234977375 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5200000.00 .082497900 5300000.00 .0861876502225 3103 14 5400000.00 .090482100 5500000.00 .095555300 5600000.00 .1014841502225 3103 15 5700000.00 .108255000 5800000.00 .115777000 5900000.00 .1239030002225 3103 16 6000000.00 .132481167 6100000.00 .141251500 6200000.00 .1500218332225 3103 17 6300000.00 .158677500 6400000.00 .167038500 6500000.00 .1749920002225 3103 18 6600000.00 .182459000 6700000.00 .189393500 6800000.00 .1957795002225 3103 19 6900000.00 .201623000 7000000.00 .206947500 7100000.00 .2116860002225 3103 20 7200000.00 .216090000 7300000.00 .220127250 7400000.00 .2237977502225 3103 21 7500000.00 .227143500 7600000.00 .230164500 7700000.00 .2331855002225 3103 22 7800000.00 .235943125 7900000.00 .238437375 8000000.00 .2409316252225 3103 23 8100000.00 .243425875 8200000.00 .245762125 8300000.00 .2479403752225 3103 24 8400000.00 .250118625 8500000.00 .252296875 8600000.00 .2544751252225 3103 25 8700000.00 .256653375 8800000.00 .258831625 8900000.00 .2610098752225 3103 26 9000000.00 .263188125 9100000.00 .265366375 9200000.00 .2675446252225 3103 27 9300000.00 .269722875 9400000.00 .271901125 9500000.00 .2740793752225 3103 28 9600000.00 .276257625 9700000.00 .278435875 9800000.00 .2804955832225 3103 29 9900000.00 .282436750 10000000.0 .284377917 10100000.0 .2863190832225 3103 30 10200000.0 .288260250 10300000.0 .290201417 10400000.0 .2919556252225 3103 31 10500000.0 .293522875 10600000.0 .295090125 10700000.0 .2966573752225 3103 32 10800000.0 .297992875 10900000.0 .299096625 11000000.0 .3002003752225 3103 33 11100000.0 .301304125 11200000.0 .302152667 11300000.0 .3027460002225 3103 34 11400000.0 .303339333 11500000.0 .303725000 11600000.0 .3039030002225 3103 35 11700000.0 .303857833 11800000.0 .303589500 11900000.0 .3033211672225 3103 36 12000000.0 .302775333 12100000.0 .301952000 12200000.0 .3011286672225 3103 37 12300000.0 .300027833 12400000.0 .298649500 12500000.0 .2972711672225 3103 38 12600000.0 .295628333 12700000.0 .293721000 12800000.0 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1.501230-4 2.033080-4 2.545010-4222533103 46 3.170430-4 3.686510-4 3.737870-4 3.635790-3 2.830400-3 1.521650-3222533103 47 5.839200-4 1.340470-4 3.008480-5 1.061840-4 2.360330-4 3.400020-4222533103 48 3.792170-4 3.475140-4 2.621310-4 1.531760-4 5.276980-5-1.371600-5222533103 49 -3.243860-5-1.615290-6 7.111340-5 1.724880-4 2.872210-4 4.005680-4222533103 50 5.355510-4 6.209800-4 5.540420-4 2.992230-3 2.210920-3 1.321870-3222533103 51 6.648290-4 2.798550-4 1.044490-4 5.950710-5 7.769920-5 1.107890-4222533103 52 1.301710-4 1.246390-4 9.638370-5 5.562990-5 1.508460-5-1.444240-5222533103 53 -2.648250-5-1.932520-5 4.843760-6 4.135800-5 8.442420-5 1.467160-4222533103 54 2.047540-4 2.065660-4 2.102090-3 1.594860-3 1.020360-3 5.333650-4222533103 55 1.840260-4-2.719890-5-1.218160-4-1.306990-4-8.748800-5-2.378860-5222533103 56 3.485020-5 7.171340-5 7.966920-5 5.982700-5 1.883050-5-3.412800-5222533103 57 -8.970580-5-1.398910-4-1.786300-4-2.044440-4-1.784490-4-7.872330-5222533103 58 1.445120-3 1.083580-3 6.775910-4 3.228050-4 6.477870-5-8.405100-5222533103 59 -1.334110-4-1.076300-4-3.862740-5 4.120460-5 1.054510-4 1.378040-4222533103 60 1.330070-4 9.502850-5 3.371470-5-3.860100-5-1.097120-4-1.690850-4222533103 61 -2.150220-4-1.835830-4-2.528830-5 9.449480-4 7.149460-4 4.759940-4222533103 62 2.760980-4 1.382640-4 6.658500-5 5.147420-5 7.484400-5 1.155630-4222533103 63 1.546680-4 1.791070-4 1.831920-4 1.679270-4 1.389590-4 1.042630-4222533103 64 7.221500-5 5.034500-5 5.195700-5 1.295790-4 3.101290-4 6.811540-4222533103 65 6.084510-4 5.221670-4 4.374450-4 3.632780-4 3.038860-4 2.601880-4222533103 66 2.310620-4 2.144760-4 2.083910-4 2.113000-4 2.224300-4 2.416450-4222533103 67 2.692310-4 3.056580-4 3.514030-4 4.379840-4 5.881100-4 7.771760-4222533103 68 6.991220-4 7.367750-4 7.216420-4 6.605220-4 5.672850-4 4.601860-4222533103 69 3.585400-4 2.790880-4 2.332240-4 2.259000-4 2.562190-4 3.191390-4222533103 70 4.074230-4 5.132960-4 6.295330-4 8.071950-4 1.037690-3 1.238670-3222533103 71 8.790310-4 9.295870-4 8.911490-4 7.816890-4 6.311210-4 4.740970-4222533103 72 3.424850-4 2.593850-4 2.363030-4 2.736960-4 3.638140-4 4.942810-4222533103 73 6.512370-4 8.215070-4 1.072590-3 1.374030-3 1.596580-3 1.030230-3222533103 74 1.017920-3 9.115610-4 7.451500-4 5.608290-4 3.988850-4 2.897880-4222533103 75 2.500630-4 2.824190-4 3.789560-4 5.254750-4 7.054170-4 9.026670-4222533103 76 1.194860-3 1.543810-3 1.794240-3 1.031340-3 9.444440-4 7.900760-4222533103 77 6.093440-4 4.433830-4 3.246260-4 2.720760-4 2.908640-4 3.750690-4222533103 78 5.118720-4 6.855040-4 8.801580-4 1.175590-3 1.539350-3 1.814440-3222533103 79 8.888590-4 7.683740-4 6.179780-4 4.727510-4 3.619820-4 3.042870-4222533103 80 3.066400-4 3.663150-4 4.741640-4 6.179180-4 7.847390-4 1.048310-3222533103 81 1.391040-3 1.675830-3 6.951350-4 5.922860-4 4.861340-4 3.987540-4222533103 82 3.458160-4 3.349860-4 3.668070-4 4.367340-4 5.373920-4 6.604250-4222533103 83 8.666360-4 1.155910-3 1.425790-3 5.437350-4 4.853040-4 4.317030-4222533103 84 3.935570-4 3.777810-4 3.872170-4 4.214350-4 4.778520-4 5.527710-4222533103 85 6.898540-4 9.026440-4 1.128670-3 4.746400-4 4.584020-4 4.431900-4222533103 86 4.334040-4 4.325120-4 4.425760-4 4.644360-4 4.980230-4 5.687730-4222533103 87 6.952440-4 8.510190-4 4.782020-4 4.903100-4 4.974660-4 5.014420-4222533103 88 5.045490-4 5.088950-4 5.162030-4 5.356930-4 5.801400-4 6.480210-4222533103 89 5.318150-4 5.641130-4 5.877260-4 6.024600-4 6.093140-4 6.096570-4222533103 90 6.009780-4 5.800660-4 5.550780-4 6.279110-4 6.821970-4 7.248890-4222533103 91 7.532320-4 7.662530-4 7.557490-4 6.950360-4 5.859180-4 7.756650-4222533103 92 8.574140-4 9.223740-4 9.650720-4 9.782880-4 9.082420-4 7.360870-4222533103 93 9.863660-4 1.097040-3 1.182050-3 1.240980-3 1.193460-3 9.889720-4222533103 94 1.260120-3 1.394240-3 1.515950-3 1.521600-3 1.321900-3 1.583350-3222533103 95 1.778520-3 1.865200-3 1.710910-3 2.097680-3 2.348220-3 2.342350-3222533103 96 2.885560-3 3.210870-3 4.018350-3 222533103 97 222533 099999 2225 0 0 0 0 0 0 0 2.20470E+4 4.65484E+1 0 0 34 102228 1451 1 0.0 0.0 0 0 0 62228 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022228 1451 3 3.00000E+2 0.0 1 0 163 72228 1451 4 22-Ti- 47 FEI/ANL EVAL-Oct95 K.I.Zolotarev, C.Philis et al. 2228 1451 5 DIST-Feb2004 2228 1451 6 ----IRDF-2002 MATERIAL 2228 2228 1451 7 -----INCIDENT NEUTRON DATA 2228 1451 8 ------ENDF-6 FORMAT 2228 1451 9 ***************************************************************** 2228 1451 10 22-TI-47 FEI EVAL-Oct95 K.I.Zolotarev 2228 1451 11 DIST-Feb02 Mod1 20020203 2228 1451 12 ----BROND-2 MATERIAL 2228 REVISION 1 2228 1451 13 -----INCIDENT NEUTRON DATA 2228 1451 14 ------ENDF-6 FORMAT 2228 1451 15 ------Russian Reactor Dosimetry File RRDF-2002 2228 1451 16 ***************************************************************** 2228 1451 17 22-TI- 47 ANL EVAL-JAN77 C.PHILIS,O.BERSILLON,D.SMITH,ETC2228 1451 18 DIST-FEB91 910201 2228 1451 19 ----IRDF-90 MATERIAL 2228 2228 1451 20 ENDF/B-VI MATERIAL 2228 2228 1451 21 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 2228 1451 22 ***************************************************************** 2228 1451 23 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2228 1451 24 DOSIMETRY 2228 1451 25 ***************************************************************** 2228 1451 26 ******** Start of (N,NP) bibliographical component ******* 2228 1451 27 ***************************************************************** 2228 1451 28 Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko 2228 1451 29 2228 1451 30 For this special purpose library it was decided the reaction 2228 1451 31 MF=3 MT=28+104 would use MF=10, MT=5 to present this data. 2228 1451 32 This was done after processing through the codes. The 2228 1451 33 co-variance file was converted from MF/MT=33/28 to MF/MT=40/5 2228 1451 34 ******************************************************************2228 1451 35 MF=3 2228 1451 36 MT= 28 -(n,np+pn+d) cross section data 2228 1451 37 -------------------------------------- 2228 1451 38 In this section is given the sum of cross section of the reac-2228 1451 39 tions Ti47(n,np)Sc46m+g , Ti47(n,pn)Sc46m+g and Ti47(n,d)Sc46m+g. 2228 1451 40 Excitation function for the Ti47(n,x)Sc46m+g reaction in the 2228 1451 41 energy region from threshold to 20 MeV was evaluated by means of 2228 1451 42 statistical analysis of experimental cross section data [1-6] and 2228 1451 43 data from STAPRE [7] calculation. 2228 1451 44 All experimental data were renormalized to the new standards 2228 1451 45 for monitor reactions cross sections and decay data. 2228 1451 46 The final procedure of evaluation Ti47(n,x)Sc46m+g excitation 2228 1451 47 function from threshold to 20 MeV has been carried out within the 2228 1451 48 framework of generalized least squares method. Rational function 2228 1451 49 was used as model function [8]. Calculations was performed by 2228 1451 50 means of Pade-2 code [9]. 2228 1451 51 U-235 thermal fission [10] and Cf-252 spontaneous fission 2228 1451 52 neutron spectra [11] averaged cross-sections calculated from the 2228 1451 53 evaluated Ti47(n,x)Sc46m+g excitation function are the following: 2228 1451 54 2228 1451 55 -------------------------------------------- 2228 1451 56 TYPE OF SPECTRUM I , mb (calc.) 2228 1451 57 --------------------------I----------------- 2228 1451 58 U-235 neutron fission I 8.1158E-3 2228 1451 59 CF-252 spontan. fission I 1.9201E-2 2228 1451 60 2228 1451 61 MF=33 2228 1451 62 MT= 28 -(n,np+pn+d) cross section cov. matrix 2228 1451 63 --------------------------------------------- 2228 1451 64 Uncertainties in the evaluated excitation function for the 2228 1451 65 reaction Ti-47(n,x)Sc-46m+g are given in the form of relative 2228 1451 66 covariance matrix for the 17-neutron energy groups (LB=5). Cova- 2228 1451 67 riance matrix of uncertainties was calculated simultaneously with 2228 1451 68 recommended cross section data by means of PADE-2 code. 2228 1451 69 Eigenvalues of the 6-th digits relative covariance matrix 2228 1451 70 given in the 33-file are the following: 2228 1451 71 2228 1451 72 3.91044E-07 4.46920E-07 5.37497E-07 7.02044E-07 2228 1451 73 9.85160E-07 1.54860E-06 2.86775E-06 6.97782E-06 2228 1451 74 2.67223E-05 1.90995E-04 9.14779E-04 5.12707E-03 2228 1451 75 8.27566E-03 1.15029E-02 1.68918E-02 5.65241E-02 2228 1451 76 8.27240E-01 2228 1451 77 2228 1451 78 References : 2228 1451 79 1. W.G.Cross, H.L.Pai Progress Rep. EANDC(CAN)-16, p.1, Jan.1963 2228 1451 80 2. H.L.Pai Canadian J. of Physics, v.44, p.2337, 1966 2228 1451 81 3. I.Ribansky, S.Gmuca J. Phys.G, v.9, p.1537, December 1983 2228 1451 82 4. N.I.Molla et al. Report INDC(BAN)-003, September 1986 2228 1451 83 5. Y.Ikeda et al. Report JAERI-1312, March 1988 2228 1451 84 6. Y.Uno et al. Report JAERI-M-93-046, p.247-256, 1993 2228 1451 85 7. M.Uhl, B.Strohmaier Computer Code STAPRE for Particle Induced 2228 1451 86 Activation Cross Section and Related Quantities, Report 2228 1451 87 IRK 76-01, Vienna, 1976 2228 1451 88 8. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2228 1451 89 st's Meeting on Evaluation and Processing of Covariance Data, 2228 1451 90 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2228 1451 91 9. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2228 1451 92 10. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2228 1451 93 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2228 1451 94 11. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2228 1451 95 ******************************************************************2228 1451 96 ******** End of (N,NP) bibliographical component ******* 2228 1451 97 ***************************************************************** 2228 1451 98 ***************************************************************** 2228 1451 99 ******** Start of (N,P) bibliographical component ******* 2228 1451 100 ***************************************************************** 2228 1451 101 ** 2228 1451 102 DOCUMENTATION ANL/NDM-27 (N,P) 2228 1451 103 ** 2228 1451 104 COMMENTS- (N,P) EVAL IN ENERGY RANGES 1.10E+06 TO 1.0E+07 AND 2228 1451 105 1.35E+07 TO 2.0E+07 BASED ON EXPERIMENTAL DATA.INTERPOLATION AND2228 1451 106 EXTRAPOLATION TO OTHER ENERGIES BASED ON THEORETICAL CALCULATION2228 1451 107 PERFORMED AT BRUYERES LE CHATEL USING CODE DEVELOPED BY JARRE. 2228 1451 108 PSEUDO-THRESHOLD=0.5 MEV PER P.C. MACFARLANE TO MAGURNO 6/20/83 2228 1451 109 ** 2228 1451 110 REFERENCES 2228 1451 111 AL61 ALLAN D. NP24,274,61 2228 1451 112 AR78 ARTHUR E. PC 1978 2228 1451 113 AR75 ARMITAGE P.C. NNDC 2228 1451 114 CR62 CROSS W.AND PAI H. EANDC(CAN)-16,2,62 2228 1451 115 FU79 C.Y.FU PRIVATE COMM 1979 2228 1451 116 GH71 GHORAI S.AND CARROLL E. NSE 43,233,71 2228 1451 117 GO62 GONZALES ET.AL. PR126,271,62 2228 1451 118 HI HILLMAN M. NP37,78,62 2228 1451 119 LE69 LEVKOVSKII V. Y.F.10,44,69 2228 1451 120 PA66 PAI H.CJP 44 2337,66 2228 1451 121 PO59 POULARIKIS A. AND FINK R. PR115,989,59 2228 1451 122 SM75 SMITH D.L.AND MEADOWS J. ANL/NDM10,75 2228 1451 123 TI68 TIKKU ET.AL. SYMP ON NUCL CROSS SECT AND TECH WASH.1968 2228 1451 124 ******************************************************************2228 1451 125 ******** End of (N,P) bibliographical component ******* 2228 1451 126 ***************************************************************** 2228 1451 127 The Q values and threshold energies were updated prior to pro- 2228 1451 128 cessing through the codes to comply with the values obtained 2228 1451 129 using the NNDC calculation program which is based on the 1995 2228 1451 130 Update to the Atomic mass Evaluation. 2228 1451 131 2228 1451 132 File 2 added to the pointwise file containing only the effective 2228 1451 133 scattering radius with no resonance parameters given. 2228 1451 134 Taken from ENDF/B-VI 2228 1451 135 2228 1451 136 Threshold of 0.0 added at 1.0E-5 ev to MF/MT=3/103 after 2228 1451 137 processing thro. the pre-procssing codes. 2228 1451 138 2228 1451 139 2228 1451 140 ***************************************************************** 2228 1451 141 ***************** Program LINEAR (VERSION 2002-1) ***************2228 1451 142 For All Data Greater than 1.0000E-10 barns in Absolute Value 2228 1451 143 Data Linearized to Within an Accuracy of .100000000 per-cent 2228 1451 144 ***************** Program SIGMA1 (VERSION 2002-1) ***************2228 1451 145 Data Doppler Broadened to 300.000000 Kelvin 2228 1451 146 for All Data Greater than 1.0000E-10 barns in Absolute Value 2228 1451 147 Data Linearized to Within an Accuracy pf .100000000 per-cent 2228 1451 148 ***************** Program FIXUP (Version 2002-1) ****************2228 1451 149 Corrected ZA/AWR in All Sections-----------------------------Yes 2228 1451 150 Corrected Thresholds-----------------------------------------Yes 2228 1451 151 Extended Cross Sections to 20 MeV----------------------------No 2228 1451 152 Allow Cross Section Deletion---------------------------------No 2228 1451 153 Allow Cross Section Reconstruction---------------------------No 2228 1451 154 Make All Cross Sections Non-Negative-------------------------Yes 2228 1451 155 Delete Energies Not in Ascending Order-----------------------Yes 2228 1451 156 Deleted Duplicate Points-------------------------------------Yes 2228 1451 157 Check for Ascending MAT/MF/MT Order--------------------------Yes 2228 1451 158 Check for Legal MF/MT Numbers--------------------------------Yes 2228 1451 159 Allow Creation of Missing Sections---------------------------No 2228 1451 160 Allow Insertion of Energy Points-----------------------------No 2228 1451 161 Create Uniform Energy Grid-----------------------------------No 2228 1451 162 Delete Section if Cross Section =0 at All Energies-----------Yes 2228 1451 163 ***************************************************************** 2228 1451 164 ***************** Program GROUPIE (VERSION 2002-1) **************2228 1451 165 Unshielded Group Averages Using 640 Groups 2228 1451 166 Weighting Spectrum: Flat (Constant) Spectrum 2228 1451 167 1 451 174 12228 1451 168 2 151 4 12228 1451 169 3 103 70 02228 1451 170 8 5 2 12228 1451 171 10 5 42 12228 1451 172 33 103 115 12228 1451 173 40 5 36 02228 1451 174 2228 1 099999 2228 0 0 0 2.20470E+4 4.65480E+1 0 0 1 02228 2151 1 2.204700+4 1.000000+0 0 0 1 02228 2151 2 1.000000-5 1.000000+5 0 0 0 02228 2151 3 3.500000+0 0.573300+0 0 0 0 02228 2151 4 2228 2 099999 2228 0 0 0 2.20470E+4 4.65484E+1 0 0 0 02228 3103 1 1.82300E+5 1.82300E+5 0 0 1 1992228 3103 2 199 1 2228 3103 3 690000.000 5.16780E-5 720000.000 .000310068 760000.000 .0006201362228 3103 4 800000.000 .000805043 840000.000 .000864789 880000.000 .0009245352228 3103 5 920000.000 .000984281 960000.000 .001044027 1000000.00 .0013463902228 3103 6 1100000.00 .001891370 1200000.00 .002436350 1300000.00 .0029813302228 3103 7 1400000.00 .003526310 1500000.00 .004697020 1600000.00 .0064934602228 3103 8 1700000.00 .008289900 1800000.00 .010086340 1900000.00 .0118827802228 3103 9 2000000.00 .013905700 2100000.00 .016155100 2200000.00 .0184045002228 3103 10 2300000.00 .020653900 2400000.00 .022903300 2500000.00 .0250525002228 3103 11 2600000.00 .027101500 2700000.00 .029097500 2800000.00 .0310405002228 3103 12 2900000.00 .032983500 3000000.00 .034727500 3100000.00 .0362725002228 3103 13 3200000.00 .037817500 3300000.00 .039362500 3400000.00 .0409075002228 3103 14 3500000.00 .042653900 3600000.00 .044601700 3700000.00 .0465495002228 3103 15 3800000.00 .048497300 3900000.00 .050445100 4000000.00 .0520001002228 3103 16 4100000.00 .053162300 4200000.00 .054324500 4300000.00 .0554867002228 3103 17 4400000.00 .056648900 4500000.00 .057830600 4600000.00 .0590318002228 3103 18 4700000.00 .060233000 4800000.00 .061434200 4900000.00 .0626354002228 3103 19 5000000.00 .064098300 5100000.00 .065822900 5200000.00 .0675475002228 3103 20 5300000.00 .069272100 5400000.00 .070996700 5500000.00 .0723986002228 3103 21 5600000.00 .073477800 5700000.00 .074557000 5800000.00 .0756362002228 3103 22 5900000.00 .076715400 6000000.00 .078082700 6100000.00 .0797381002228 3103 23 6200000.00 .081393500 6300000.00 .083048900 6400000.00 .0847043002228 3103 24 6500000.00 .086108400 6600000.00 .087261200 6700000.00 .0884140002228 3103 25 6800000.00 .089566800 6900000.00 .090719600 7000000.00 .0919561002228 3103 26 7100000.00 .093276300 7200000.00 .094596500 7300000.00 .0959167002228 3103 27 7400000.00 .097236900 7500000.00 .098933300 7600000.00 .1010059002228 3103 28 7700000.00 .103078500 7800000.00 .105151100 7900000.00 .1072237002228 3103 29 8000000.00 .109065000 8100000.00 .110675000 8200000.00 .1122850002228 3103 30 8300000.00 .113895000 8400000.00 .115505000 8500000.00 .1168730002228 3103 31 8600000.00 .117999000 8700000.00 .119125000 8800000.00 .1202510002228 3103 32 8900000.00 .121377000 9000000.00 .122315000 9100000.00 .1230650002228 3103 33 9200000.00 .123815000 9300000.00 .124565000 9400000.00 .1253150002228 3103 34 9500000.00 .126237000 9600000.00 .127331000 9700000.00 .1284250002228 3103 35 9800000.00 .129519000 9900000.00 .130613000 10000000.0 .1312550002228 3103 36 10100000.0 .131445000 10200000.0 .131635000 10300000.0 .1318250002228 3103 37 10400000.0 .132015000 10500000.0 .132205000 10600000.0 .1323950002228 3103 38 10700000.0 .132585000 10800000.0 .132775000 10900000.0 .1329650002228 3103 39 11000000.0 .132870000 11100000.0 .132490000 11200000.0 .1321100002228 3103 40 11300000.0 .131730000 11400000.0 .131350000 11500000.0 .1309700002228 3103 41 11600000.0 .130590000 11700000.0 .130210000 11800000.0 .1298300002228 3103 42 11900000.0 .129450000 12000000.0 .128879750 12100000.0 .1281192502228 3103 43 12200000.0 .127358750 12300000.0 .126598250 12400000.0 .1258377502228 3103 44 12500000.0 .125077250 12600000.0 .124316750 12700000.0 .1235562502228 3103 45 12800000.0 .122795750 12900000.0 .122035250 13000000.0 .1212747502228 3103 46 13100000.0 .120514250 13200000.0 .119753750 13300000.0 .1189932502228 3103 47 13400000.0 .118232750 13500000.0 .117472250 13600000.0 .1167117502228 3103 48 13700000.0 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1.600000+7 1.650000+7222840 5 6 1.700000+7 1.750000+7 1.800000+7 1.850000+7 1.900000+7 1.950000+7222840 5 7 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222840 5 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222840 5 9 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0222840 5 10 0.000000+0 8.224320-1 3.047800-2-5.099880-3 2.660400-2 2.567630-2222840 5 11 8.789060-3-2.089150-4 4.143110-4 5.627230-3 1.120560-2 1.502500-2222840 5 12 1.653970-2 1.606880-2 1.428180-2 1.189650-2 9.535380-3 7.674830-3222840 5 13 2.317220-2 1.579900-2 4.911180-3 9.215290-4 7.415560-4 1.566370-3222840 5 14 2.456920-3 3.187640-3 3.730820-3 4.106900-3 4.345820-3 4.476950-3222840 5 15 4.526420-3 4.516220-3 4.464290-3 4.384770-3 2.410370-2 1.179110-2222840 5 16 2.085650-3-4.356060-5 1.197690-3 2.968140-3 4.280220-3 4.991600-3222840 5 17 5.234730-3 5.189020-3 5.005500-3 4.790790-3 4.611590-3 4.504530-3222840 5 18 4.485670-3 1.255610-2 7.443600-3 3.314840-3 1.501210-3 1.219560-3222840 5 19 1.623630-3 2.199950-3 2.702100-3 3.042030-3 3.212750-3 3.242780-3222840 5 20 3.171400-3 3.036720-3 2.870740-3 7.555650-3 4.699770-3 2.094550-3222840 5 21 5.902510-4 3.201110-5 5.900340-5 3.656710-4 7.479170-4 1.087880-3222840 5 22 1.328070-3 1.448730-3 1.451680-3 1.349620-3 3.780760-3 2.278690-3222840 5 23 1.087240-3 3.782550-4 7.219620-5 3.330810-5 1.404120-4 3.044130-4222840 5 24 4.669730-4 5.938050-4 6.675780-4 6.820470-4 2.094230-3 1.740670-3222840 5 25 1.376410-3 1.067600-3 8.374700-4 6.850910-4 5.990970-4 5.649730-4222840 5 26 5.687140-4 5.983170-4 6.442110-4 2.137400-3 2.230550-3 2.111580-3222840 5 27 1.877100-3 1.606180-3 1.352240-3 1.146690-3 1.004800-3 9.311090-4222840 5 28 9.235670-4 2.710740-3 2.837360-3 2.718770-3 2.461500-3 2.152640-3222840 5 29 1.853540-3 1.602850-3 1.421840-3 1.319530-3 3.190670-3 3.236060-3222840 5 30 3.076280-3 2.802720-3 2.488110-3 2.183810-3 1.922800-3 1.723960-3222840 5 31 3.447240-3 3.420070-3 3.236110-3 2.964070-3 2.657910-3 2.356140-3222840 5 32 2.084160-3 3.528090-3 3.454820-3 3.259670-3 2.991250-3 2.688170-3222840 5 33 2.378610-3 3.491600-3 3.386500-3 3.181570-3 2.911070-3 2.602730-3222840 5 34 3.371280-3 3.242790-3 3.031150-3 2.760710-3 3.195830-3 3.061290-3222840 5 35 2.860660-3 3.016960-3 2.911930-3 2.923980-3 222840 5 36 222840 099999 2228 0 0 0 0 0 0 0 2.20480E+4 4.75360E+1 0 0 34 102231 1451 1 0.0 0.0 0 0 0 62231 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022231 1451 3 3.00000E+2 0.0 1 0 242 72231 1451 4 22-Ti- 48 FEI EVAL-Feb02 K.I.Zolotarev 2231 1451 5 DIST-Feb2004 2231 1451 6 ----IRDF-2002 MATERIAL 2231 2231 1451 7 -----INCIDENT NEUTRON DATA 2231 1451 8 ------ENDF-6 FORMAT 2231 1451 9 ***************************************************************** 2231 1451 10 ********************** SPECIAL PURPOSE FILE ***************** 2231 1451 11 DOSIMETRY 2231 1451 12 2231 1451 13 For this special purpose library it was decided the reaction 2231 1451 14 MF=3 MT=28+104 would use MF=10, MT=5 to present this data. 2231 1451 15 This was done after processing through the codes. The 2231 1451 16 co-variance file was converted from MF/MT=33/28 to MF/MT=40/5 2231 1451 17 ***************************************************************** 2231 1451 18 22-TI-48 FEI EVAL-Feb02 K.I.Zolotarev 2231 1451 19 DIST-Feb02 20020214 2231 1451 20 ----BROND-2 MATERIAL 2231 2231 1451 21 -----INCIDENT NEUTRON DATA 2231 1451 22 ------ENDF-6 FORMAT 2231 1451 23 ------Russian Reactor Dosimetry File RRDF-2002 2231 1451 24 ***************************************************************** 2231 1451 25 Author of evaluation: K.I.Zolotarev 2231 1451 26 ***************************************************************** 2231 1451 27 MF= 3 2231 1451 28 MT= 28 -(n,np+pn+d) cross section data 2231 1451 29 -------------------------------------- 2231 1451 30 In this section is given the sum of cross sections of the re- 2231 1451 31 actions Ti-48(n,np)Sc-47 , Ti-48(n,pn)Sc-47 and Ti-48(n,d)Sc-47. 2231 1451 32 Excitation function for the Ti-48(n,x)Sc-47 reaction in the 2231 1451 33 energy region from threshold to 20 MeV was evaluated by means of 2231 1451 34 statistical analysis of experimental cross section data [1-12] 2231 1451 35 and data from STAPRE [13] calculation. 2231 1451 36 Analised experimental data were corrected to the new stan- 2231 1451 37 dards for monitor reactions cross sections and decay data. 2231 1451 38 The final procedure of evaluation Ti-48(n,x)Sc-47 excitation 2231 1451 39 function from threshold to 20 MeV has been carried out within the 2231 1451 40 framework of generalized least squares method. Rational function 2231 1451 41 was used as model function [14]. Calculations was performed by 2231 1451 42 means of Pade-2 code [15]. 2231 1451 43 U-235 thermal fission [16] and Cf-252 spontanious fission 2231 1451 44 neutron spectra [17] averaged cross-sections calculated from the 2231 1451 45 evaluated Ti-48(n,x)Sc-47 excitation function are the following: 2231 1451 46 2231 1451 47 -------------------------------------------- 2231 1451 48 TYPE OF SPECTRUM I , mb (calc.) 2231 1451 49 --------------------------I----------------- 2231 1451 50 U-235 neutron fission I 1.6558E-3 2231 1451 51 CF-252 spontan. fission I 4.2891E-3 2231 1451 52 2231 1451 53 MT=103 -(n,p) cross section 2231 1451 54 ------------------------------------- 2231 1451 55 Excitation function for the Ti-48(n,p)Sc-48 reaction in the 2231 1451 56 energy region from threshold to 20 MeV was evaluated by means of 2231 1451 57 of statistical analysis of experimental cross section data [1-2], 2231 1451 58 [4], [6-11], [18-41] and data from STAPRE [13] calculation. 2231 1451 59 The energy dependence of cross-section from 4.7 MeV to thres- 2231 1451 60 hold was extrapolated with L=0 penetrability function for the 2231 1451 61 the outgoing p + Sc48 channal [42]. 2231 1451 62 Analised microscopic experimental data were renormalized to 2231 1451 63 the new recommended standards for monitor reaction cross sections 2231 1451 64 and decay data. All experiments performed with Ti-natural samples 2231 1451 65 were corrected for the contribution of the Ti-49(n,np+pn)Sc-48 2231 1451 66 and Ti-49(n,d)Sc-48 reactions. Total cross section for these reac-2231 1451 67 tions were taken from ref. [43]. Experimental data from ref. [24] 2231 1451 68 were used partially. It was taken into account only data measured 2231 1451 69 relative Fe-56(n,p)Mn-56 monitor reaction. Data of Firkin [6] 2231 1451 70 obtained in the experiment with neutrons from D(d,n)He3 reaction 2231 1451 71 were renormalized to the results of his measurements with 14.1 2231 1451 72 MeV neutrons from T(d,n)He4 reaction. Cross section data of Smith 2231 1451 73 and Meadows [26] measured in the energy range 5.964 - 9.952 MeV 2231 1451 74 with using D(d,n)He3 neutron source were renormalized to the 2231 1451 75 value 30.7 mb at 10 MeV [38]. The correction factors for the 2231 1451 76 experimental data [6] and [26] were Fc=0.91853 and Fc=1.11111 , 2231 1451 77 respectively 2231 1451 78 Experimental cross section data [44-51] were rejected due to 2231 1451 79 their discrepancy with the main bulk of experimental data [1-2], 2231 1451 80 [4], [6-11], [18-41] and data from theoretical model calculatiod. 2231 1451 81 Statistical analysis of input cross section data was carried 2231 1451 82 out by means of PADE-2 code [15]. Rational function was used as 2231 1451 83 the model function [14]. 2231 1451 84 U-235 thermal fission [16] and Cf-252 spontanious fission 2231 1451 85 neutron spectra [17] averaged cross-sections calculated from the 2231 1451 86 evaluated Ti-48(n,p)Sc-48 excitation function are the following: 2231 1451 87 2231 1451 88 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2231 1451 89 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2231 1451 90 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2231 1451 91 U-235 neutron fission ³ 0.30430 ³ 0.305 +- 0.020 [52] 2231 1451 92 ³ ³ 0.302 +- 0.010 [53] 2231 1451 93 ³ ³ 0.3007+- 0.0054 [54] 2231 1451 94 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2231 1451 95 Cf-252 spont. fission ³ 0.42629 ³ 0.4275+- 0.0078 [55] 2231 1451 96 ³ ³ 0.4247+- 0.0080 [56] 2231 1451 97 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2231 1451 98 2231 1451 99 MF=33 2231 1451 100 MT= 28 -(n,np+pn+d) cross section cov. matrix 2231 1451 101 --------------------------------------------- 2231 1451 102 Uncertainties in the evaluated excitation function for the 2231 1451 103 reaction Ti-48(n,x)Sc-47 are given in the form of relative cova- 2231 1451 104 riance matrix for the 15-neutron energy groups (LB=5). Covariance 2231 1451 105 matrix of uncertainties was calculated simultaneously with 2231 1451 106 recommended cross section data by means of PADE-2 code. 2231 1451 107 Eigenvalues of the 6-th digits relative covariance matrix 2231 1451 108 given in the 33-file are the following: 2231 1451 109 2231 1451 110 3.25031E-07 3.96552E-07 7.16579E-07 1.55391E-06 2231 1451 111 4.04176E-06 1.39260E-05 8.49003E-05 2.17726E-03 2231 1451 112 8.18710E-03 1.48647E-02 2.37004E-02 3.37652E-02 2231 1451 113 4.90298E-02 9.66009E-02 5.46438E-01 2231 1451 114 2231 1451 115 MT=103 -(n,p) cross section cov. matrix 2231 1451 116 --------------------------------------------- 2231 1451 117 Uncertainties in the evaluated excitation function for the 2231 1451 118 reaction Ti-48(n,p)Sc-48 are given in the form of relative cova- 2231 1451 119 riance matrix for the 26-neutron energy groups (LB=5). Covariance 2231 1451 120 matrix of uncertainties was calculated simultaneously with 2231 1451 121 recommended cross section data by means of PADE-2 code. 2231 1451 122 Eigenvalues of the 6-th digits relative covariance matrix 2231 1451 123 given in the 33-file are the following: 2231 1451 124 2231 1451 125 5.39341E-09 5.73858E-09 5.82529E-09 6.42044E-09 2231 1451 126 6.80480E-09 7.28300E-09 7.51108E-09 8.50118E-09 2231 1451 127 9.53827E-09 1.09813E-08 1.24867E-08 1.39806E-08 2231 1451 128 1.87772E-08 2.72244E-08 4.05820E-08 6.49966E-08 2231 1451 129 1.23779E-07 3.87183E-07 6.43873E-06 1.17950E-03 2231 1451 130 6.38433E-03 8.03793E-03 1.04937E-02 1.43313E-02 2231 1451 131 2.41167E-02 3.79706E-02 2231 1451 132 2231 1451 133 References : 2231 1451 134 1. W.G.Cross, H.L.Pai Prog. Report EANDC(CAN)-16, p.1, Jan. 1963 2231 1451 135 2. H.L.Pai Canadian J. of Physics, v.44, p.2337, 1966 2231 1451 136 3. S.M.Grimes et al. Nucl. Sci. Eng., v.62, p.187, Feb. 1977 2231 1451 137 4. S.M.Qaim, N.I.Molla Nucl. Phys., v.A283, p.269, June 1977 2231 1451 138 5. M.Viennot et al. Progress Report MOH-5, p.10, 1982 2231 1451 139 6. S.Firkin Report AERE-M-3350, Harwell, September 1983 2231 1451 140 7. I.Ribansky, S.Gmuca J. Phys.G, v.9, p.1537, December 1983 2231 1451 141 8. N.I.Molla et al. Report INDC(BAN)-003, September 1986 2231 1451 142 9. Y.Ikeda et al. Report JAERI-1312, March 1988 2231 1451 143 10. W.V.Hecker et al. Nucl. Inst. Meth., Sec.B, v.40/41, p.478, 2231 1451 144 April 1989 2231 1451 145 11. M.Viennot et al. Nucl. Sci. Eng., v.108, p.289, July 1991 2231 1451 146 12. Y.Uno et al. Report JAERI-M-93-046, p.247-256, 1993 2231 1451 147 13. M.Uhl, B.Strohmaier Computer Code STAPRE for Particle Induced 2231 1451 148 Activation Cross Section and Related Quantities, Report 2231 1451 149 IRK 76-01, Vienna, 1976 2231 1451 150 14. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2231 1451 151 st's Meeting on Evaluation and Processing of Covariance Data, 2231 1451 152 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2231 1451 153 15. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2231 1451 154 16. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2231 1451 155 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2231 1451 156 17. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2231 1451 157 18. A.Poularikas, R.W.Fink Physical Review, v.115, p.989, 1959 2231 1451 158 19. F.Gabbard, B.D.Kern Physical Review, v.128, p.1276, 1962 2231 1451 159 20. M.Hillman Nucl. Phys., v.37, p.78, 1962 2231 1451 160 21. M.Bormann et al. Nucl. Phys., v.63, p.438, March 1965 2231 1451 161 22. V.N.Levkovskij et al. Yadernaja Fizika (Sov.), v.10, n.1, 2231 1451 162 p.44, July 1969 2231 1451 163 23. S.K.Ghorai et al. J. Nucl. Energy, v.25, p.319, August 1971 2231 1451 164 24. Y.Lukic, E.E.Carroll Nucl. Sci. Eng., v.43, p.233, 1971 2231 1451 165 25. J.Prasad, D.C.Sarcar Nuovo Cimento, v.A3, p.467, June 1971 2231 1451 166 26. D.L.Smith,J.W.Meadows Nucl. Sci. Eng., v.58, p.314, Nov. 1975 2231 1451 167 27. W.Mannhart, H.Vonach Zeitsch. fur Physik, Sect.A, v.272, 2231 1451 168 p.279, March 1975 2231 1451 169 28. K.Kayashima e.a. Prog. Report NEANDC(J)-61U, p.94, Sep. 1979 2231 1451 170 29. M.Viennot e.a. Proc. of Int. Conf.on Nuclear Data for Science 2231 1451 171 and Technology, Antwerpen, Belgium, 6-10 Sep. 1982, p.406 2231 1451 172 30. L.R.Greenwood Report DOE-ER-0046-21, p.15, May 1985 2231 1451 173 31. R.Pepelnik et al. Progress Report NEANDC(E)-262U, (5), p.32, 2231 1451 174 June 1985 2231 1451 175 32. J.P.Gupta et al. Pramana, v.24, p.637, 1985 2231 1451 176 33. Hoang Dac Luc et al. Progress Report INDC(VN)-5, Sep. 1986 2231 1451 177 34. K.Kobayashi, I.Kimura Proc. of an Intern. Conf. on Nuclear 2231 1451 178 Data for Science and Technology, 30 May - 3 June 1988, Mito, 2231 1451 179 Japan, Saikon Publishing Co., LTD, pp.261-265, 1989 2231 1451 180 35. Lu Han-Lin et al. Report INDC(CPR)-16, August 1989 2231 1451 181 36. Y.Ikeda et al. Report JAERI-M-91-032, p.281, March 1991 2231 1451 182 37. M.Viennot et al. Nucl. Sci. Eng., v.108, p.289, July 1991 2231 1451 183 38. S.M.Qaim et al. Proc. of Int. Conference on Nuclear Data for 2231 1451 184 Science and Technology, Julich, FRG, 13-17 May 1991. Springer 2231 1451 185 Verlag, Berlin - Heidelberg, 1992, p.297-300 2231 1451 186 39. I.Garlea et al. J. Rev. Roum. Phys., v.37, n.1, p.19, 1992 2231 1451 187 40. Yuan Junqian et al. High Energy Physics and Nuclear Physics 2231 1451 188 (China), v.16, n.1, p.57, January 1992 2231 1451 189 41. A.Grallert et al. Progress Report,INDC(NDS)-286, p.131, 1993 2231 1451 190 42. S.A.Badikov, A.B.Pashchenko Voprosy Atomnoy Nauki i Tekhniki 2231 1451 191 Ser. Jadernye Konstanty, 2(53), p.70, 1987 2231 1451 192 43. K.Zolotarev RRDF-2002, MAT=2243, eval. February 2002 2231 1451 193 44. E.B.Paul, R.L.Clarke Canadian J. Phys., v.31, p.267, 1953 2231 1451 194 45. D.L.Allan Nucl. Phys., v.24, p.274, April 1961 2231 1451 195 46. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 2231 1451 196 47. H.K.Vonach et al. Proc. of 2-nd Conference on Nuclear Cross- 2231 1451 197 Sections and Technology, Washington D.C., 4-7 March 1968, 2231 1451 198 v.2, p.885 2231 1451 199 48. D.Crumpton J. Inorg. Nucl. Chem., v.31, p.3727, Dec. 1969 2231 1451 200 49. V.K.Tikku et al. Proc. of Nucl. Phys. and Solid State Phys. 2231 1451 201 Symp., Chandigarh, v.2, p.115, December 1972 2231 1451 202 50. R.Spangler et al. J. Trans. Amer. Nucl. Soc., v.22, p.818, 2231 1451 203 November 1975 2231 1451 204 51. K.T.Osman, F.I.Habbani Report, INDC(SUD)-001, October 1996 2231 1451 205 52. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2231 1451 206 Washington D.C., 25-28 April 1989, v.2, p.923 2231 1451 207 53. W.Mannhart Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry, 2231 1451 208 Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985 2231 1451 209 54. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 2231 1451 210 55. W.Mannhart Handbook on Nuclear Activation Cross Sections , 2231 1451 211 IAEA Tech. Report Ser. No.273, p.413, 1987 2231 1451 212 56. W.Mannhart Validation of Differential Cross Sections with 2231 1451 213 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2231 1451 214 September 2002 2231 1451 215 ***************************************************************** 2231 1451 216 File 2 added to the pointwise file containing only the effective 2231 1451 217 scattering radius with no resonance parameters given. 2231 1451 218 Taken from ENDF/B-VI 2231 1451 219 ***************************************************************** 2231 1451 220 2231 1451 221 ***************** Program LINEAR (VERSION 2002-1) ***************2231 1451 222 For All Data Greater than 1.0000E-10 barns in Absolute Value 2231 1451 223 Data Linearized to Within an Accuracy of .100000000 per-cent 2231 1451 224 ***************** Program SIGMA1 (VERSION 2002-1) ***************2231 1451 225 Data Doppler Broadened to 300.000000 Kelvin 2231 1451 226 for All Data Greater than 1.0000E-10 barns in Absolute Value 2231 1451 227 Data Linearized to Within an Accuracy pf .100000000 per-cent 2231 1451 228 ***************** Program FIXUP (Version 2002-1) ****************2231 1451 229 Corrected ZA/AWR in All Sections-----------------------------Yes 2231 1451 230 Corrected Thresholds-----------------------------------------Yes 2231 1451 231 Extended Cross Sections to 20 MeV----------------------------No 2231 1451 232 Allow Cross Section Deletion---------------------------------No 2231 1451 233 Allow Cross Section Reconstruction---------------------------No 2231 1451 234 Make All Cross Sections Non-Negative-------------------------Yes 2231 1451 235 Delete Energies Not in Ascending Order-----------------------Yes 2231 1451 236 Deleted Duplicate Points-------------------------------------Yes 2231 1451 237 Check for Ascending MAT/MF/MT Order--------------------------Yes 2231 1451 238 Check for Legal MF/MT Numbers--------------------------------Yes 2231 1451 239 Allow Creation of Missing Sections---------------------------No 2231 1451 240 Allow Insertion of Energy Points-----------------------------No 2231 1451 241 Create Uniform Energy Grid-----------------------------------No 2231 1451 242 Delete Section if Cross Section =0 at All Energies-----------Yes 2231 1451 243 ***************** Program GROUPIE (VERSION 2002-1) **************2231 1451 244 Unshielded Group Averages Using 640 Groups 2231 1451 245 Weighting Spectrum: Flat (Constant) Spectrum 2231 1451 246 1 451 253 12231 1451 247 2 151 4 12231 1451 248 3 103 60 12231 1451 249 8 5 2 12231 1451 250 10 5 39 12231 1451 251 33 103 71 12231 1451 252 40 5 30 12231 1451 253 2231 1 099999 2231 0 0 0 2.20480E+4 4.75360E+1 0 0 1 02231 2151 1 2.204800+4 1.000000+0 0 0 1 02231 2151 2 1.000000-5 1.000000-5 0 0 0 02231 2151 3 0.000000+0 0.573300+0 0 0 0 02231 2151 4 2231 2 099999 2231 0 0 0 2.20480E+4 4.75360E+1 0 0 0 02231 3103 1 -3.21180E+6-3.21180E+6 0 0 1 1692231 3103 2 169 1 2231 3103 3 3200000.00 4.44705E-9 3300000.00 1.46148E-7 3400000.00 3.52717E-72231 3103 4 3500000.00 5.59287E-7 3600000.00 7.65856E-7 3700000.00 9.72426E-72231 3103 5 3800000.00 1.17900E-6 3900000.00 1.75487E-6 4000000.00 2.60258E-62231 3103 6 4100000.00 3.40719E-6 4200000.00 4.52734E-6 4300000.00 6.44311E-62231 3103 7 4400000.00 9.78106E-6 4500000.00 1.53460E-5 4600000.00 2.41598E-52231 3103 8 4700000.00 3.75024E-5 4800000.00 5.69529E-5 4900000.00 8.44353E-52231 3103 9 5000000.00 .000122266 5100000.00 .000173193 5200000.00 .0002404202231 3103 10 5300000.00 .000327615 5400000.00 .000438894 5500000.00 .0005787552231 3103 11 5600000.00 .000751970 5700000.00 .000963413 5800000.00 .0012178252231 3103 12 5900000.00 .001519525 6000000.00 .001872085 6100000.00 .0022779952231 3103 13 6200000.00 .002738340 6300000.00 .003252590 6400000.00 .0038184952231 3103 14 6500000.00 .004432120 6600000.00 .005088075 6700000.00 .0057798702231 3103 15 6800000.00 .006500350 6900000.00 .007244910 7000000.00 .0080010502231 3103 16 7100000.00 .008760743 7200000.00 .009523988 7300000.00 .0102872332231 3103 17 7400000.00 .011050478 7500000.00 .011802500 7600000.00 .0125433002231 3103 18 7700000.00 .013284100 7800000.00 .014013106 7900000.00 .0147303172231 3103 19 8000000.00 .015447528 8100000.00 .016164739 8200000.00 .0168819502231 3103 20 8300000.00 .017599161 8400000.00 .018316372 8500000.00 .0190335832231 3103 21 8600000.00 .019750794 8700000.00 .020481567 8800000.00 .0212259002231 3103 22 8900000.00 .021970233 9000000.00 .022732733 9100000.00 .0235134002231 3103 23 9200000.00 .024294067 9300000.00 .025097267 9400000.00 .0259230002231 3103 24 9500000.00 .026748733 9600000.00 .027599500 9700000.00 .0284753002231 3103 25 9800000.00 .029351100 9900000.00 .030252433 10000000.0 .0311793002231 3103 26 10100000.0 .032106167 10200000.0 .033060613 10300000.0 .0340426382231 3103 27 10400000.0 .035024663 10500000.0 .036006688 10600000.0 .0370165502231 3103 28 10700000.0 .038054250 10800000.0 .039091950 10900000.0 .0401296502231 3103 29 11000000.0 .041167350 11100000.0 .042205050 11200000.0 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1.205820-3 1.175750-3 5.878370-4-9.490980-5223140 5 18 -5.712350-4-7.367790-4-6.212280-4 8.396540-3 7.438880-3 3.738740-3223140 5 19 1.061670-3 2.405110-4 6.122910-4 1.371180-3 2.005320-3 2.316480-3223140 5 20 2.311780-3 2.087370-3 1.621120-2 1.515040-2 8.632830-3 3.121490-3223140 5 21 1.138420-3 2.178660-3 4.619640-3 6.965430-3 8.348450-3 8.521370-3223140 5 22 1.895660-2 1.478520-2 8.727250-3 4.770970-3 3.793980-3 4.917860-3223140 5 23 6.814000-3 8.444710-3 9.280860-3 1.554920-2 1.270370-2 8.999380-3223140 5 24 6.138550-3 4.681970-3 4.463460-3 5.026080-3 5.915220-3 1.336970-2223140 5 25 1.145390-2 8.468040-3 5.709430-3 3.904690-3 3.230420-3 3.514540-3223140 5 26 1.161810-2 1.015980-2 8.083180-3 6.198510-3 4.939490-3 4.404970-3223140 5 27 1.095260-2 1.089590-2 1.022320-2 9.212560-3 8.099470-3 1.315310-2223140 5 28 1.425810-2 1.409830-2 1.289180-2 1.700900-2 1.794020-2 1.714270-2223140 5 29 1.989720-2 1.988000-2 2.083550-2 223140 5 30 223140 099999 2231 0 0 0 0 0 0 0 2.20490E+4 4.85274E+1 0 0 34 202234 1451 1 0.0 0.0 0 0 0 62234 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022234 1451 3 3.00000E+2 0.0 1 0 116 52234 1451 4 22-Ti- 49 FEI EVAL-Feb02 K.I.Zolotarev 2234 1451 5 DIST-Feb2004 2234 1451 6 ----IRDF-2002 MATERIAL 2234 2234 1451 7 -----INCIDENT NEUTRON DATA 2234 1451 8 ------ENDF-6 FORMAT 2234 1451 9 ***************************************************************** 2234 1451 10 ********************** SPECIAL PURPOSE FILE ***************** 2234 1451 11 DOSIMETRY 2234 1451 12 2234 1451 13 For this special purpose library it was decided the reaction 2234 1451 14 MF=3 MT=28+104 would use MF=10, MT=5 to present this data. 2234 1451 15 This was done after processing through the codes. The 2234 1451 16 co-variance file was converted from MF/MT=33/28 to MF/MT=40/5 2234 1451 17 ***************************************************************** 2234 1451 18 22-TI-49 FEI EVAL-Feb02 K.I.Zolotarev 2234 1451 19 DIST-Feb02 2234 1451 20 ----BROND-2 MATERIAL 2234 2234 1451 21 -----INCIDENT NEUTRON DATA 2234 1451 22 ------ENDF-6 FORMAT 2234 1451 23 ------Russian Reactor Dosimetry File RRDF-2002 2234 1451 24 ***************************************************************** 2234 1451 25 Author of evaluation: K.I.Zolotarev 2234 1451 26 ***************************************************************** 2234 1451 27 MF= 3 2234 1451 28 MT= 28 -(n,np+pn+d) cross section data 2234 1451 29 -------------------------------------- 2234 1451 30 In this section is given the sum of cross section of the reac-2234 1451 31 tions Ti-49(n,np)Sc-48 , Ti-49(n,pn)Sc-48 and Ti-49(n,d)Sc-48. 2234 1451 32 Excitation function for the Ti-49(n,x)Sc-48 reaction in the 2234 1451 33 energy region from threshold to 20 MeV was evaluated by means of 2234 1451 34 statistical analysis of experimental cross section data [1-5] and 2234 1451 35 data from STAPRE [6] calculation. 2234 1451 36 Analysed experimental data were renormalized to the new stan- 2234 1451 37 dards for monitor reactions cross sections and decay data. Data 2234 1451 38 of Pai [2] measured in the energy region 16-19.5 MeV with using 2234 1451 39 Van de Graaff accelerator were renormalized to the results of 2234 1451 40 the theoretical model calculation. 2234 1451 41 The final procedure of evaluation Ti-49(n,x)Sc-48 excitation 2234 1451 42 function from threshold to 20 MeV has been carried out within the 2234 1451 43 framework of generalized least squares method. Rational function 2234 1451 44 was used as model function [7]. Calculations was performed by 2234 1451 45 means of Pade-2 code [8]. 2234 1451 46 U-235 thermal fission [9] and Cf-252 spontaneous fission 2234 1451 47 neutron spectra [10] averaged cross-sections calculated from the 2234 1451 48 evaluated Ti-49(n,x)Sc-48 excitation function are the following: 2234 1451 49 2234 1451 50 -------------------------------------------- 2234 1451 51 TYPE OF SPECTRUM I , mb (calc.) 2234 1451 52 --------------------------I----------------- 2234 1451 53 U-235 neutron fission I 1.0041E-3 2234 1451 54 CF-252 spontan. fission I 2.6070E-3 2234 1451 55 2234 1451 56 MF=33 2234 1451 57 MT= 28 -(n,np+pn+d) cross section cov. matrix 2234 1451 58 --------------------------------------------- 2234 1451 59 Uncertainties in the evaluated excitation function for the 2234 1451 60 reaction Ti-49(n,x)Sc-48 are given in the form of relative cova- 2234 1451 61 riance matrix for the 14-neutron energy groups (LB=5). Covariance 2234 1451 62 matrix of uncertainties was calculated simultaneously with 2234 1451 63 recommended cross section data by means of PADE-2 code. 2234 1451 64 Eigenvalues of the 6-th digits relative covariance matrix 2234 1451 65 given in the 33-file are the following: 2234 1451 66 2234 1451 67 2.34704E-08 3.44912E-08 5.30971E-08 9.47934E-08 2234 1451 68 1.99158E-07 5.01449E-07 1.49592E-06 5.55617E-06 2234 1451 69 1.56887E-04 2.23251E-03 8.77261E-03 1.50447E-02 2234 1451 70 3.12585E-02 8.36148E-02 2234 1451 71 2234 1451 72 References : 2234 1451 73 1. W.G.Cross, H.L.Pai Progress Report, EANDC(CAN)-16, p.1, 2234 1451 74 January 1963 2234 1451 75 2. H.L.Pai Canadian J. of Physics, v.44, p.2337, 1966 2234 1451 76 3. S.M.Qaim Nucl. Phys., v.A382, n.2, p.255, July 1982 2234 1451 77 4. I.Ribansky, S.Gmuca J. Phys.G, v.9, p.1537, December 1983 2234 1451 78 5. Y.Ikeda et al. Report JAERI-1312, March 1988 2234 1451 79 6. M.Uhl, B.Strohmaier Computer Code STAPRE for Particle Induced 2234 1451 80 Activation Cross Section and Related Quantities, Report 2234 1451 81 IRK 76-01, Vienna, 1976 2234 1451 82 7. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2234 1451 83 st's Meeting on Evaluation and Processing of Covariance Data, 2234 1451 84 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2234 1451 85 8. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2234 1451 86 9. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2234 1451 87 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2234 1451 88 10. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2234 1451 89 ***************************************************************** 2234 1451 90 File 2 added to the pointwise file containing only the effective 2234 1451 91 scattering radius with no resonance parameters given. 2234 1451 92 Taken from JENDL-3.2 2234 1451 93 ***************************************************************** 2234 1451 94 2234 1451 95 ***************** Program LINEAR (VERSION 2002-1) ***************2234 1451 96 For All Data Greater than 1.0000E-10 barns in Absolute Value 2234 1451 97 Data Linearized to Within an Accuracy of .100000000 per-cent 2234 1451 98 ***************** Program SIGMA1 (VERSION 2002-1) ***************2234 1451 99 Data Doppler Broadened to 300.000000 Kelvin 2234 1451 100 for All Data Greater than 1.0000E-10 barns in Absolute Value 2234 1451 101 Data Linearized to Within an Accuracy pf .100000000 per-cent 2234 1451 102 ***************** Program FIXUP (Version 2002-1) ****************2234 1451 103 Corrected ZA/AWR in All Sections-----------------------------Yes 2234 1451 104 Corrected Thresholds-----------------------------------------Yes 2234 1451 105 Extended Cross Sections to 20 MeV----------------------------No 2234 1451 106 Allow Cross Section Deletion---------------------------------No 2234 1451 107 Allow Cross Section Reconstruction---------------------------No 2234 1451 108 Make All Cross Sections Non-Negative-------------------------Yes 2234 1451 109 Delete Energies Not in Ascending Order-----------------------Yes 2234 1451 110 Deleted Duplicate Points-------------------------------------Yes 2234 1451 111 Check for Ascending MAT/MF/MT Order--------------------------Yes 2234 1451 112 Check for Legal MF/MT Numbers--------------------------------Yes 2234 1451 113 Allow Creation of Missing Sections---------------------------No 2234 1451 114 Allow Insertion of Energy Points-----------------------------No 2234 1451 115 Create Uniform Energy Grid-----------------------------------No 2234 1451 116 Delete Section if Cross Section =0 at All Energies-----------Yes 2234 1451 117 ***************** Program GROUPIE (VERSION 2002-1) **************2234 1451 118 Unshielded Group Averages Using 640 Groups 2234 1451 119 Weighting Spectrum: Flat (Constant) Spectrum 2234 1451 120 1 451 125 12234 1451 121 2 151 4 12234 1451 122 8 5 2 12234 1451 123 10 5 39 12234 1451 124 40 5 27 12234 1451 125 2234 1 099999 2234 0 0 0 2.20490E+4 4.85274E+1 0 0 1 02234 2151 1 2.20490+ 4 1.00000+ 0 0 0 1 02234 2151 2 1.00000- 5 1.00000+ 5 0 0 0 02234 2151 3 3.50000+ 0 4.50000- 1 0 0 0 02234 2151 4 2234 2 099999 2234 0 0 0 2.20490E+4 4.85274E+1 0 0 1 12234 8 5 1 2.104800+4 0.000000+0 10 0 0 02234 8 5 2 2234 8 099999 2234 0 0 0 2.20490E+4 4.85274E+1 0 0 1 0223410 5 1 -9.12956E+6-9.12956E+6 21048 0 1 108223410 5 2 108 1 223410 5 3 9300000.00 1.10601E-9 9400000.00 4.31560E-9 9500000.00 7.57585E-9223410 5 4 9600000.00 1.08361E-8 9700000.00 1.40964E-8 9800000.00 1.73566E-8223410 5 5 9900000.00 2.06169E-8 10000000.0 4.97413E-8 10100000.0 1.04730E-7223410 5 6 10200000.0 1.59719E-7 10300000.0 2.14707E-7 10400000.0 2.69696E-7223410 5 7 10500000.0 7.67187E-7 10600000.0 1.70718E-6 10700000.0 2.64718E-6223410 5 8 10800000.0 3.58717E-6 10900000.0 4.52716E-6 11000000.0 6.74413E-6223410 5 9 11100000.0 1.09550E-5 11200000.0 1.67516E-5 11300000.0 2.44527E-5223410 5 10 11400000.0 3.44093E-5 11500000.0 4.70069E-5 11600000.0 6.26681E-5223410 5 11 11700000.0 8.18556E-5 11800000.0 .000105075 11900000.0 .000132876223410 5 12 12000000.0 .000165858 12100000.0 .000204674 12200000.0 .000250028223410 5 13 12300000.0 .000302684 12400000.0 .000363468 12500000.0 .000433267223410 5 14 12600000.0 .000513036 12700000.0 .000603801 12800000.0 .000706656223410 5 15 12900000.0 .000822770 13000000.0 .000953387 13100000.0 .001099825223410 5 16 13200000.0 .001263485 13300000.0 .001445825 13400000.0 .001648390223410 5 17 13500000.0 .001872795 13600000.0 .002120700 13700000.0 .002393845223410 5 18 13800000.0 .002694020 13900000.0 .003023020 14000000.0 .003382710223410 5 19 14100000.0 .003774980 14200000.0 .004201675 14300000.0 .004664635223410 5 20 14400000.0 .005165665 14500000.0 .005706495 14600000.0 .006288770223410 5 21 14700000.0 .006914025 14800000.0 .007583640 14900000.0 .008298825223410 5 22 15000000.0 .009060585 15100000.0 .009869710 15200000.0 .010726650223410 5 23 15300000.0 .011631650 15400000.0 .012584700 15500000.0 .013585300223410 5 24 15600000.0 .014632650 15700000.0 .015725600 15800000.0 .016862750223410 5 25 15900000.0 .018042250 16000000.0 .019270825 16100000.0 .020527875223410 5 26 16200000.0 .021818075 16300000.0 .023141425 16400000.0 .024496200223410 5 27 16500000.0 .025882400 16600000.0 .027268600 16700000.0 .028676978223410 5 28 16800000.0 .030107533 16900000.0 .031538089 17000000.0 .032968644223410 5 29 17100000.0 .034399200 17200000.0 .035829756 17300000.0 .037260311223410 5 30 17400000.0 .038690867 17500000.0 .040121422 17600000.0 .041516575223410 5 31 17700000.0 .042876325 17800000.0 .044236075 17900000.0 .045595825223410 5 32 18000000.0 .046914233 18100000.0 .048191300 18200000.0 .049468367223410 5 33 18300000.0 .050704483 18400000.0 .051899650 18500000.0 .053094817223410 5 34 18600000.0 .054247567 18700000.0 .055357900 18800000.0 .056468233223410 5 35 18900000.0 .057530588 19000000.0 .058544963 19100000.0 .059559338223410 5 36 19200000.0 .060573713 19300000.0 .061538125 19400000.0 .062452575223410 5 37 19500000.0 .063367025 19600000.0 .064281475 19700000.0 .065157833223410 5 38 19800000.0 .065996100 19900000.0 .066834367 20000000.0 0.0 223410 5 39 223410 099999 2234 0 0 0 2.20490E+4 4.85274E+1 0 0 1 0223440 5 1 -9.12956E+6-9.12956E+6 0 1 0 1223440 5 2 1.000000+1 1.000000+0 0 5 0 1223440 5 3 0.000000+0 0.000000+0 1 5 136 16223440 5 4 1.000000-5 9.300000+6 1.350000+7 1.400000+7 1.450000+7 1.500000+7223440 5 5 1.550000+7 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7223440 5 6 1.850000+7 1.900000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0223440 5 7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223440 5 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0223440 5 9 0.000000+0 2.340470-2 1.573760-2 1.165090-2 8.156660-3 5.541340-3223440 5 10 3.969430-3 3.457650-3 3.824990-3 4.757050-3 5.897230-3 6.966960-3223440 5 11 7.813930-3 8.420550-3 8.858660-3 1.315090-2 1.087230-2 8.474760-3223440 5 12 6.164580-3 4.207140-3 2.837340-3 2.197190-3 2.273200-3 2.909620-3223440 5 13 3.862550-3 4.881200-3 5.757010-3 6.348110-3 9.682370-3 8.120140-3223440 5 14 6.337310-3 4.558730-3 3.043450-3 2.025410-3 1.624510-3 1.820480-3223440 5 15 2.467110-3 3.358720-3 4.283100-3 5.063930-3 7.358580-3 6.232520-3223440 5 16 4.886070-3 3.524410-3 2.383810-3 1.652690-3 1.421190-3 1.664730-3223440 5 17 2.277620-3 3.115020-3 4.034000-3 5.813390-3 5.091770-3 4.162120-3223440 5 18 3.178860-3 2.326160-3 1.761320-3 1.580970-3 1.811250-3 2.427040-3223440 5 19 3.374180-3 5.065940-3 4.757790-3 4.186930-3 3.465820-3 2.759470-3223440 5 20 2.250120-3 2.087010-3 2.377510-3 3.180650-3 5.131690-3 5.149110-3223440 5 21 4.804660-3 4.208600-3 3.567080-3 3.106570-3 3.039490-3 3.527780-3223440 5 22 5.806850-3 6.020090-3 5.794240-3 5.284890-3 4.721960-3 4.359860-3223440 5 23 4.419540-3 6.857770-3 7.193930-3 7.072420-3 6.651570-3 6.153380-3223440 5 24 5.798070-3 8.165580-3 8.611450-3 8.566600-3 8.159510-3 7.554320-3223440 5 25 9.678720-3 1.017690-2 1.011670-2 9.575680-3 1.129210-2 1.182470-2223440 5 26 1.177030-2 1.316420-2 1.408870-2 1.651520-2 223440 5 27 223440 099999 2234 0 0 0 0 0 0 0 2.30510E+4 5.05063E+1 0 0 34 102328 1451 1 0.0 0.0 0 0 0 62328 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022328 1451 3 3.00000E+2 0.0 1 0 185 42328 1451 4 23-V - 51 FEI EVAL-Dec01 K.I.Zolotarev 2328 1451 5 DIST-Feb2004 2328 1451 6 ----IRDF-2002 MATERIAL 2328 2328 1451 7 -----INCIDENT NEUTRON DATA 2328 1451 8 ------ENDF-6 FORMAT 2328 1451 9 ***************************************************************** 2328 1451 10 23-V-51 FEI EVAL-Dec01 K.I.Zolotarev 2328 1451 11 DIST-Jan02 20020118 2328 1451 12 ----BROND-2 MATERIAL 2328 2328 1451 13 -----INCIDENT NEUTRON DATA 2328 1451 14 ------ENDF-6 FORMAT 2328 1451 15 ------Russian Reactor Dosimetry File RRDF-2002 2328 1451 16 ***************************************************************** 2328 1451 17 Author of evaluation: K.I.Zolotarev 2328 1451 18 ***************************************************************** 2328 1451 19 MF=3 2328 1451 20 MT=107 -(n,a) cross section 2328 1451 21 ------------------------------------- 2328 1451 22 Excitation function for the V-51(n,a)Sc-48 reaction in the 2328 1451 23 energy region from threshold to 20 MeV was evaluated by means of 2328 1451 24 statistical analysis of experimental cross section data [1-21] 2328 1451 25 and data from GNASH calculation [45]. 2328 1451 26 All analysed microscopic experimental data [1-38] and integ- 2328 1451 27 ral experimental data [39-44] were renormalized to the new recom- 2328 1451 28 mended standards for monitor reactions cross sections and decay 2328 1451 29 data. 2328 1451 30 Experimental cross section data from ref. [22-38] were rejec- 2328 1451 31 ted due to their big discrepancy with the main bulk of experimen- 2328 1451 32 al data [1-21], data from theoretical model calculation and data 2328 1451 33 from (n,a) cross section systematics. 2328 1451 34 Statistical analysis of input cross section data was carried 2328 1451 35 out by means of PADE-2 code [46]. Rational function was used as 2328 1451 36 the model function [47]. 2328 1451 37 The evaluated V-51(n,a)Sc-48 excitation function averaged 2328 1451 38 on U-235 neutron fission spectrum [48] and Cf-252 spontaneous 2328 1451 39 fission neutron spectrum [49] gives the next values : 2328 1451 40 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2328 1451 41 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2328 1451 42 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2328 1451 43 U-235 neutron fission ³ 0.024414 ³ 0.0235 +-0.0015 [40] 2328 1451 44 ³ ³ 0.0215 +-0.0013 [41] 2328 1451 45 ³ ³ 0.02438+-0.00056 [42] 2328 1451 46 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2328 1451 47 CF-252 spont. fission ³ 0.038514 ³ 0.03904+-0.00086 [43] 2328 1451 48 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2328 1451 49 2328 1451 50 MF=33 2328 1451 51 MT=107 -(n,a) cross section cov. matrix 2328 1451 52 --------------------------------------- 2328 1451 53 Uncertainties in the evaluated excitation function for the 2328 1451 54 reaction V-51(n,a)Sc-48 are given in the form of relative covari- 2328 1451 55 ance matrix for the 26-neutron energy groups (LB=5). Covariance 2328 1451 56 matrix of uncertainties was calculated simultaneously with recom- 2328 1451 57 mended cross section data by means of PADE-2 code. 2328 1451 58 Eigenvalues of the 6-th digits relative covariance matrix 2328 1451 59 given in the 33-file are the following: 2328 1451 60 2328 1451 61 4.46067E-09 4.97872E-09 5.09809E-09 5.50951E-09 2328 1451 62 6.22582E-09 6.26808E-09 6.80068E-09 7.58613E-09 2328 1451 63 9.63723E-09 1.29149E-08 1.82432E-08 2.72742E-08 2328 1451 64 4.31460E-08 7.37989E-08 1.54144E-07 3.75108E-07 2328 1451 65 2.13791E-06 1.13740E-04 6.78127E-04 9.37900E-04 2328 1451 66 1.40076E-03 2.54577E-03 4.01102E-03 5.22789E-03 2328 1451 67 1.00172E-02 1.34609E-02 2328 1451 68 2328 1451 69 References : 2328 1451 70 1. H.Vonach, H.Muenzer Oesterr. Akad. Wiss., Math + Naturw. 2328 1451 71 Anzeiger, v.95, p.199, November 1958 2328 1451 72 2. H.K.Vonach et al. Proc. 2nd Conf. on Nuclear Cross-Sections 2328 1451 73 and Technology, Washington D.C., 4-7 March 1968, v.2, p.885 2328 1451 74 3. V.N.Levkovskiy et al. Yadernaja Fizika (Sov.), v.10, n.1, 2328 1451 75 p.44, July 1968 2328 1451 76 4. J.C.Robertson,B.Audric,P.Kolkowski J. of Nucl. Energy, v.27, 2328 1451 77 p.531, August 1973 2328 1451 78 5. A.Paulsen et al. Atomkernenergie, v.22, p.291, Sep. 1974 2328 1451 79 6. W.Mannhart, H.Vonach Zeitschrift fuer Physik, Section A, 2328 1451 80 v.272, p.279, March 1975 2328 1451 81 7. S.M.Qaim, N.I.Molla Proc. of 9th Symposium on Fusion Techno- 2328 1451 82 logy, Garmisch, 14-18 June 1976, FRG, Pergamon Press, p.589 2328 1451 83 8. O.Schwerer et al. Oesterr. Akad. Wiss., Math + Naturw. 2328 1451 84 Anzeiger, v.113, p.153, June 1976 2328 1451 85 9. E.Zupranska et al. Progress Report INR-1821/I/PL/A, 1980 2328 1451 86 E.Zupranska et al. Acta. Phys. Pol., v.B11, p.853, Nov. 1980 2328 1451 87 10. I.Kanno, J.W.Meadows, D.L.Smith Annals of Nuclear Energy, 2328 1451 88 v.11, p.623, December 1984 2328 1451 89 11. J.W.Meadows, D.L.Smith et al. Annals of Nuclear Energy, v.14, 2328 1451 90 p.489, September 1987 2328 1451 91 12. Y.Ikeda et al. Report JAERI-1312, March 1988 2328 1451 92 13. K.Kobayashi, I.Kimura Proc. of an Intern. Conf. on Nuclear 2328 1451 93 Data for Science and Technology, 30 May - 3 June 1988, Mito, 2328 1451 94 Japan, Saikon Publishing Co., LTD, pp.261-265, 1989 ; 2328 1451 95 14. Lu Hanlin et al. Report INDC(CPR)-16, August 1989 2328 1451 96 15. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 2328 1451 97 16. N.I.Molla et al. Proc. of International Conference on Nuclear 2328 1451 98 Data for Science and Technology, Gatlinburg, Tennessee, USA, 2328 1451 99 9-13 May 1994, v.2, pp.938-940 2328 1451 100 17. A.A.Filatenkov et al. VANT, Ser.: Yadernye Konstanty, v.2, 2328 1451 101 p.8, Moscow, 1996 2328 1451 102 18. D.Subasi et al. Nucl. Sci. Eng., to be published in 1997 2328 1451 103 19. A.D.Majdeddin "Measurement and analysis of excitation func- 2328 1451 104 tions for fast neutron induced reactions from threshold to 2328 1451 105 20 MeV". PHD dissertation, Inst. of Experimental Physics , 2328 1451 106 Kossuth University, Hungary, June 1998 2328 1451 107 20. A.A.Filatenkov et al. Report RI-252, St.Petersburg, May 1999 2328 1451 108 21. P.Reimer et al. Progress Report On Nuclear Data Research in 2328 1451 109 the FRG, INDC(Ger)-047, July 2001 2328 1451 110 22. I.Kumabe J. Phys. Soc. Jap., v.13, p.325, 1958 2328 1451 111 23. E.B.Paul, R.L.Clarke Canadian J. of Phys., v.31, p.267, 1953 2328 1451 112 24. A.Poularikas, R.W.Fink Progress Report A-ARK-60, p.3, 1960 2328 1451 113 25. F.Gabbard et al. Bull. Amer. Phys. Soc., v.5, p.42(C8), 1960 2328 1451 114 26. M.Bormann et al. Phys. Rev., v.22, p.602, October 1961 2328 1451 115 27. M.Hillman Nucl. Phys., v.37, p.78, 1962 2328 1451 116 28. E.T.Bramlitt, R.W.Fink Phys. Rev., v.131, p.2649, Sep. 1963 2328 1451 117 29. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 2328 1451 118 30. D.Crumpton J. Inorg. Nucl. Chem., v.31, p.3727, Dec. 1969 2328 1451 119 31. R.Spangler et al. Transactions of the American Nuclear 2328 1451 120 Society, v.22, p.818, November 1975 2328 1451 121 32. U.Garuska et al. Progress Report INDC(POL)-9, p.16, Sep. 1978 2328 1451 122 33. W.H.Warren, W.L.Alford Annals of Nucl. Energy, v.9,p.369,1982 2328 1451 123 34. G.Helfer et al. Czechoslovak Journal of Physics, Part B, 2328 1451 124 v.34, p.30, 1984 2328 1451 125 35. R.Pepelnik et al. Int. Conf. on Nuclear Data for Basic and 2328 1451 126 Applied Science, Santa Fe, N.M., 13-17 May 1985, v.1, p.211 2328 1451 127 36. N.I.Molla et al. Progress Report INDC(BAN)-003 , Sep. 1986 2328 1451 128 37. Y.Ikeda et al. Progress Report INDC(JPN)-162/U,p.24,Aug. 1992 2328 1451 129 38. J.Cezar Suita et al. Nucl. Sci. Eng., v.126, p.101, 1997 2328 1451 130 39. F.Nasyrov,B.D.Sciborskij Atomnaya Energija (Sov.), v.25, n.5,2328 1451 131 p.437, November 1968 2328 1451 132 40. I.Kimura, K.Kobayashi, T.Shibata Nucl. Sci. Technology., v.8,2328 1451 133 p.59, February 1971 2328 1451 134 41. K.Kobayashi, I.Kimura et al. Nucl. Sci. Technology., v.13, 2328 1451 135 p.531, October 1976 2328 1451 136 42. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 2328 1451 137 43. W.Mannhart Proc. of 5-th ASTM-EURATOM Symposium on Reactor 2328 1451 138 Dosimetry. Geesthacht, Sep.24-28, 1984, Vol.2, p.813, 1985 2328 1451 139 44. J.Csikai, Z.Dezso Proc. of 4th All Union Conf. on Neutron 2328 1451 140 Physics, Kiev, 18-22 April 1977, v.3, p.32 2328 1451 141 45. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 2328 1451 142 Model Code for Calculation of Cross Section and Emission 2328 1451 143 Spectra. Report LA-6947, Los Alamos, 1977 ; 2328 1451 144 E.L.Trykiv, G.Ya.Tertychnyi Private communication, IPPE, 2328 1451 145 Obninsk, May 1999 2328 1451 146 46. S.A.Badikov et.al. Preprint FEI-1686, Obninsk, 1985 2328 1451 147 47. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2328 1451 148 st's Meeting on Evaluation and Processing of Covariance Data, 2328 1451 149 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2328 1451 150 48. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2328 1451 151 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2328 1451 152 49. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2328 1451 153 ***************************************************************** 2328 1451 154 The Q values and threshold energies were updated prior to pro- 2328 1451 155 cessing through the codes to comply with the values obtained 2328 1451 156 using the NNDC calculation program which is based on the 1995 2328 1451 157 Update to the Atomic mass Evaluation. 2328 1451 158 2328 1451 159 File 2 added to the pointwise file containing only the effective 2328 1451 160 scattering radius with no resonance parameters given. 2328 1451 161 Taken from ENDF/B-VI 2328 1451 162 ***************************************************************** 2328 1451 163 2328 1451 164 ***************** Program LINEAR (VERSION 2002-1) ***************2328 1451 165 For All Data Greater than 1.0000E-10 barns in Absolute Value 2328 1451 166 Data Linearized to Within an Accuracy of .100000000 per-cent 2328 1451 167 ***************** Program SIGMA1 (VERSION 2002-1) ***************2328 1451 168 Data Doppler Broadened to 300.000000 Kelvin 2328 1451 169 for All Data Greater than 1.0000E-10 barns in Absolute Value 2328 1451 170 Data Linearized to Within an Accuracy pf .100000000 per-cent 2328 1451 171 ***************** Program FIXUP (Version 2002-1) ****************2328 1451 172 Corrected ZA/AWR in All Sections-----------------------------Yes 2328 1451 173 Corrected Thresholds-----------------------------------------Yes 2328 1451 174 Extended Cross Sections to 20 MeV----------------------------No 2328 1451 175 Allow Cross Section Deletion---------------------------------No 2328 1451 176 Allow Cross Section Reconstruction---------------------------No 2328 1451 177 Make All Cross Sections Non-Negative-------------------------Yes 2328 1451 178 Delete Energies Not in Ascending Order-----------------------Yes 2328 1451 179 Deleted Duplicate Points-------------------------------------Yes 2328 1451 180 Check for Ascending MAT/MF/MT Order--------------------------Yes 2328 1451 181 Check for Legal MF/MT Numbers--------------------------------Yes 2328 1451 182 Allow Creation of Missing Sections---------------------------No 2328 1451 183 Allow Insertion of Energy Points-----------------------------No 2328 1451 184 Create Uniform Energy Grid-----------------------------------No 2328 1451 185 Delete Section if Cross Section =0 at All Energies-----------Yes 2328 1451 186 ***************** Program GROUPIE (VERSION 2002-1) **************2328 1451 187 Unshielded Group Averages Using 640 Groups 2328 1451 188 Weighting Spectrum: Flat (Constant) Spectrum 2328 1451 189 1 451 193 12328 1451 190 2 151 4 12328 1451 191 3 107 64 12328 1451 192 33 107 71 12328 1451 193 2328 1 099999 2328 0 0 0 2.30510E+4 5.05063E+1 0 0 1 02328 2151 1 2.30510+ 4 1.00000+ 0 0 0 1 02328 2151 2 1.00000- 5 9.80000+ 5 0 0 0 02328 2151 3 0.00000+ 0 5.44480- 1 0 0 0 02328 2151 4 2328 2 099999 2328 0 0 0 2.30510E+4 5.05063E+1 0 0 0 02328 3107 1 -2.05800E+6-2.05800E+6 0 0 1 1812328 3107 2 181 1 2328 3107 3 2000000.00 1.8941E-13 2100000.00 1.23745E-9 2200000.00 3.65187E-92328 3107 4 2300000.00 6.06629E-9 2400000.00 8.48070E-9 2500000.00 1.08951E-82328 3107 5 2600000.00 1.33095E-8 2700000.00 1.57240E-8 2800000.00 1.81384E-82328 3107 6 2900000.00 2.05528E-8 3000000.00 2.17904E-8 3100000.00 2.18512E-82328 3107 7 3200000.00 2.27955E-8 3300000.00 2.90406E-8 3400000.00 3.61691E-82328 3107 8 3500000.00 4.62735E-8 3600000.00 5.93538E-8 3700000.00 7.29986E-82328 3107 9 3800000.00 9.00308E-8 3900000.00 1.07628E-7 4000000.00 1.27030E-72328 3107 10 4100000.00 1.48239E-7 4200000.00 1.70077E-7 4300000.00 1.95692E-72328 3107 11 4400000.00 2.21935E-7 4500000.00 2.54196E-7 4600000.00 2.92475E-72328 3107 12 4700000.00 3.34447E-7 4800000.00 3.98587E-7 4900000.00 4.66421E-72328 3107 13 5000000.00 5.51411E-7 5100000.00 6.72556E-7 5200000.00 8.29793E-72328 3107 14 5300000.00 1.03209E-6 5400000.00 1.29020E-6 5500000.00 1.61995E-62328 3107 15 5600000.00 2.05269E-6 5700000.00 2.65886E-6 5800000.00 3.58291E-62328 3107 16 5900000.00 5.05849E-6 6000000.00 7.35895E-6 6100000.00 1.07116E-52328 3107 17 6200000.00 1.52802E-5 6300000.00 2.12329E-5 6400000.00 2.88160E-52328 3107 18 6500000.00 3.83819E-5 6600000.00 5.03943E-5 6700000.00 6.54289E-52328 3107 19 6800000.00 8.41770E-5 6900000.00 .000107451 7000000.00 .0001361842328 3107 20 7100000.00 .000171432 7200000.00 .000214354 7300000.00 .0002661842328 3107 21 7400000.00 .000328186 7500000.00 .000401585 7600000.00 .0004874772328 3107 22 7700000.00 .000586733 7800000.00 .000699892 7900000.00 .0008270702328 3107 23 8000000.00 .000967893 8100000.00 .001121482 8200000.00 .0012870882328 3107 24 8300000.00 .001461652 8400000.00 .001644565 8500000.00 .0018322112328 3107 25 8600000.00 .002024416 8700000.00 .002217381 8800000.00 .0024103462328 3107 26 8900000.00 .002603311 9000000.00 .002796276 9100000.00 .0029885792328 3107 27 9200000.00 .003176905 9300000.00 .003364567 9400000.00 .0035522292328 3107 28 9500000.00 .003739891 9600000.00 .003927553 9700000.00 .0041152152328 3107 29 9800000.00 .004302877 9900000.00 .004490539 10000000.0 .0046821802328 3107 30 10100000.0 .004877800 10200000.0 .005073420 10300000.0 .0052744232328 3107 31 10400000.0 .005480810 10500000.0 .005687197 10600000.0 .0059001172328 3107 32 10700000.0 .006119570 10800000.0 .006339023 10900000.0 .0065655302328 3107 33 11000000.0 .006799090 11100000.0 .007032650 11200000.0 .0072732432328 3107 34 11300000.0 .007520870 11400000.0 .007768497 11500000.0 .0080236842328 3107 35 11600000.0 .008286431 11700000.0 .008549179 11800000.0 .0088119262328 3107 36 11900000.0 .009082530 12000000.0 .009360990 12100000.0 .0096394502328 3107 37 12200000.0 .009917910 12300000.0 .010196370 12400000.0 .0104803252328 3107 38 12500000.0 .010769775 12600000.0 .011059225 12700000.0 .0113486752328 3107 39 12800000.0 .011638125 12900000.0 .011927575 13000000.0 .0122170252328 3107 40 13100000.0 .012506475 13200000.0 .012795925 13300000.0 .0130853752328 3107 41 13400000.0 .013374825 13500000.0 .013664275 13600000.0 .0139537252328 3107 42 13700000.0 .014243175 13800000.0 .014524710 13900000.0 .0147983302328 3107 43 14000000.0 .015071950 14100000.0 .015345570 14200000.0 .0156191902328 3107 44 14300000.0 .015882675 14400000.0 .016136025 14500000.0 .0163893752328 3107 45 14600000.0 .016642725 14700000.0 .016883825 14800000.0 .0171126752328 3107 46 14900000.0 .017341525 15000000.0 .017570375 15100000.0 .0177839752328 3107 47 15200000.0 .017982325 15300000.0 .018180675 15400000.0 .0183790252328 3107 48 15500000.0 .018561667 15600000.0 .018728600 15700000.0 .0188955332328 3107 49 15800000.0 .019047283 15900000.0 .019183850 16000000.0 .0193204172328 3107 50 16100000.0 .019440483 16200000.0 .019544050 16300000.0 .0196476172328 3107 51 16400000.0 .019733583 16500000.0 .019801950 16600000.0 .0198703172328 3107 52 16700000.0 .019920333 16800000.0 .019952000 16900000.0 .0199836672328 3107 53 17000000.0 .020002900 17100000.0 .019997250 17200000.0 .0199791502328 3107 54 17300000.0 .019961050 17400000.0 .019924583 17500000.0 .0198697502328 3107 55 17600000.0 .019814917 17700000.0 .019742550 17800000.0 .0196526502328 3107 56 17900000.0 .019562750 18000000.0 .019456550 18100000.0 .0193340502328 3107 57 18200000.0 .019211550 18300000.0 .019072275 18400000.0 .0189162252328 3107 58 18500000.0 .018760175 18600000.0 .018604125 18700000.0 .0184321882328 3107 59 18800000.0 .018244363 18900000.0 .018056538 19000000.0 .0178687132328 3107 60 19100000.0 .017667850 19200000.0 .017453950 19300000.0 .0172400502328 3107 61 19400000.0 .017026150 19500000.0 .016812250 19600000.0 .0165901382328 3107 62 19700000.0 .016359813 19800000.0 .016129488 19900000.0 .0158991632328 3107 63 20000000.0 0.0 2328 3107 64 2328 3 099999 2328 0 0 0 2.30510E+4 5.05063E+1 0 0 0 1232833107 1 0.000000+0 0.000000+0 0 107 0 1232833107 2 0.000000+0 0.000000+0 1 5 406 28232833107 3 1.000000-5 2.000000+6 8.000000+6 8.400000+6 8.700000+6 9.000000+6232833107 4 9.500000+6 1.000000+7 1.050000+7 1.100000+7 1.150000+7 1.200000+7232833107 5 1.250000+7 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7232833107 6 1.550000+7 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7232833107 7 1.850000+7 1.900000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0232833107 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0232833107 9 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0232833107 10 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0232833107 11 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0232833107 12 0.000000+0 6.409350-3 3.702780-3 2.209040-3 1.369340-3 8.918800-4232833107 13 8.003970-4 8.169160-4 7.473430-4 5.866800-4 3.888360-4 2.073490-4232833107 14 7.510980-5 1.598180-6-2.190230-5-1.394860-5 5.409240-6 2.066510-5232833107 15 2.397450-5 1.522380-5-1.221820-8-1.386740-5-1.964710-5-1.453750-5232833107 16 -7.426460-7 1.490920-5 2.259660-5 3.533640-3 2.721880-3 1.873850-3232833107 17 9.033920-4 2.002510-4-4.511890-5-4.519550-5 3.326440-5 9.822120-5232833107 18 1.156290-4 8.888230-5 4.058370-5-2.615250-6-2.090600-5-7.174720-6232833107 19 3.241110-5 8.182950-5 1.212640-4 1.338890-4 1.117260-4 5.896180-5232833107 20 -8.043630-6-6.393620-5-7.953580-5-2.720720-5 2.644550-3 2.251630-3232833107 21 1.492470-3 6.422470-4 9.351260-5-1.598030-4-2.144940-4-1.676780-4232833107 22 -9.073660-5-2.572730-5 1.052820-5 1.916870-5 1.098420-5-1.171280-6232833107 23 -7.662610-6-4.699890-6 5.742270-6 1.784780-5 2.499430-5 2.270910-5232833107 24 1.078710-5-6.112150-6-1.926930-5-1.772970-5 2.304190-3 1.965740-3232833107 25 1.277390-3 6.122250-4 1.405760-4-1.157460-4-1.994410-4-1.732450-4232833107 26 -9.826500-5-2.250730-5 2.462440-5 3.284700-5 7.600960-6-3.572180-5232833107 27 -7.831570-5-1.038630-4-1.029510-4-7.528640-5-2.953690-5 1.902460-5232833107 28 5.216820-5 5.264690-5 7.166370-6 2.238470-3 2.038110-3 1.502260-3232833107 29 9.134390-4 4.326360-4 1.190360-4-3.307590-5-6.563490-5-3.419560-5232833107 30 9.842040-6 3.107050-5 1.522200-5-3.221090-5-9.198690-5-1.397690-4232833107 31 -1.550570-4-1.284310-4-6.535050-5 1.421750-5 7.994730-5 9.754510-5232833107 32 3.514570-5 2.460990-3 2.328050-3 1.869570-3 1.307140-3 7.891170-4232833107 33 3.931920-4 1.413020-4 1.669140-5-1.855550-5-7.470460-6 1.409750-5232833107 34 2.418390-5 1.614900-5-4.410620-6-2.528220-5-3.424260-5-2.458510-5232833107 35 1.594710-6 3.318330-5 5.235960-5 3.840850-5 2.626390-3 2.461640-3232833107 36 2.007150-3 1.432960-3 8.739120-4 4.203870-4 1.171690-4-3.187220-5232833107 37 -5.377050-5 6.628290-6 9.978150-5 1.825280-4 2.255580-4 2.171680-4232833107 38 1.629420-4 8.216970-5 1.748080-6-5.049080-5-5.165130-5 1.281430-5232833107 39 2.588360-3 2.326970-3 1.818850-3 1.216390-3 6.537300-4 2.270770-4232833107 40 -1.614330-5-7.743920-5 2.088440-6 1.574000-4 3.175160-4 4.222880-4232833107 41 4.355210-4 3.520330-4 1.976880-4 2.239200-5-1.115150-4-1.424160-4232833107 42 -1.996670-5 2.254280-3 1.879440-3 1.339290-3 7.778230-4 3.154790-4232833107 43 2.737830-5-6.557210-5 7.354640-6 1.811410-4 3.751200-4 5.142460-4232833107 44 5.475840-4 4.604080-4 2.775360-4 5.721090-5-1.221850-4-1.787450-4232833107 45 -4.157720-5 1.655150-3 1.246880-3 7.787640-4 3.656390-4 8.818820-5232833107 46 -2.033960-5 2.532950-5 1.731710-4 3.514140-4 4.887460-4 5.329630-4232833107 47 4.643000-4 3.005900-4 9.316360-5-8.481210-5-1.539790-4-4.322840-5232833107 48 9.979500-4 6.779530-4 3.736110-4 1.516640-4 4.613630-5 5.514980-5232833107 49 1.468580-4 2.713890-4 3.757350-4 4.181980-4 3.794610-4 2.677670-4232833107 50 1.171590-4-2.033930-5-8.602870-5-2.551960-5 5.174630-4 3.454470-4232833107 51 2.033880-4 1.170720-4 9.297450-5 1.198800-4 1.743340-4 2.283430-4232833107 52 2.573910-4 2.471310-4 1.971890-4 1.213170-4 4.406540-5-4.780210-6232833107 53 3.934470-6 2.935610-4 2.326390-4 1.760670-4 1.331670-4 1.080550-4232833107 54 9.974160-5 1.033330-4 1.118310-4 1.181840-4 1.170650-4 1.060490-4232833107 55 8.591960-5 6.017890-5 3.400260-5 2.346650-4 2.112160-4 1.696470-4232833107 56 1.206830-4 7.576690-5 4.437710-5 3.183570-5 3.806620-5 5.754900-5232833107 57 8.052540-5 9.505690-5 8.939060-5 5.388830-5 2.174930-4 1.971300-4232833107 58 1.582960-4 1.124850-4 7.147940-5 4.440740-5 3.552820-5 4.323420-5232833107 59 6.047870-5 7.643890-5 7.888610-5 5.647360-5 2.120480-4 2.129620-4232833107 60 2.006700-4 1.778160-4 1.483680-4 1.168520-4 8.758620-5 6.397400-5232833107 61 4.807710-5 4.049150-5 4.054580-5 2.707820-4 3.156970-4 3.338090-4232833107 62 3.171470-4 2.661770-4 1.903260-4 1.063600-4 3.512920-5-2.393020-6232833107 63 1.149190-5 4.248520-4 4.961890-4 5.070980-4 4.505250-4 3.375260-4232833107 64 1.953650-4 6.192050-5-2.193140-5-1.942250-5 6.197670-4 6.691760-4232833107 65 6.282040-4 5.043390-4 3.275690-4 1.439350-4 5.926440-6-3.705100-5232833107 66 7.612450-4 7.584940-4 6.595730-4 4.869100-4 2.815200-4 9.398340-5232833107 67 -2.521280-5 8.108710-4 7.715490-4 6.460520-4 4.571060-4 2.389510-4232833107 68 3.009620-5 8.139380-4 7.703780-4 6.367250-4 4.204600-4 1.378650-4232833107 69 8.239680-4 7.757420-4 6.043510-4 3.003020-4 8.270000-4 7.491350-4232833107 70 5.126110-4 8.137850-4 7.643050-4 1.041490-3 232833107 71 232833 099999 2328 0 0 0 0 0 0 0 2.40520E+4 5.14943E+1 0 0 34 102431 1451 1 0.0 0.0 0 0 0 62431 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022431 1451 3 3.00000E+2 0.0 1 0 38 42431 1451 4 24-Cr- 52 IRK-VIENNA EVAL-APR90 2431 1451 5 DIST-Feb2004 2431 1451 6 ----IRDF-2002 MATERIAL 2431 2431 1451 7 -----INCIDENT NEUTRON DATA 2431 1451 8 ------ENDF-6 FORMAT 2431 1451 9 *****************************************************************2431 1451 10 DIST-JUN90 910201 2431 1451 11 IRK-EVAL.NLIB 25 2431 2431 1451 12 *****************************************************************2431 1451 13 The Q values and threshold energies were updated prior to pro- 2431 1451 14 cessing through the codes to comply with the values obtained 2431 1451 15 using the NNDC calculation program which is based on the 1995 2431 1451 16 Update to the Atomic mass Evaluation. 2431 1451 17 ***************** Program LINEAR (VERSION 2002-1) ***************2431 1451 18 For All Data Greater than 1.0000E-10 barns in Absolute Value 2431 1451 19 Data Linearized to Within an Accuracy of .100000000 per-cent 2431 1451 20 ***************** Program SIGMA1 (VERSION 2002-1) ***************2431 1451 21 Data Doppler Broadened to 300.000000 Kelvin 2431 1451 22 for All Data Greater than 1.0000E-10 barns in Absolute Value 2431 1451 23 Data Linearized to Within an Accuracy pf .100000000 per-cent 2431 1451 24 ***************** Program FIXUP (Version 2002-1) ****************2431 1451 25 Corrected ZA/AWR in All Sections-----------------------------Yes 2431 1451 26 Corrected Thresholds-----------------------------------------Yes 2431 1451 27 Extended Cross Sections to 20 MeV----------------------------No 2431 1451 28 Allow Cross Section Deletion---------------------------------No 2431 1451 29 Allow Cross Section Reconstruction---------------------------No 2431 1451 30 Make All Cross Sections Non-Negative-------------------------Yes 2431 1451 31 Delete Energies Not in Ascending Order-----------------------Yes 2431 1451 32 Deleted Duplicate Points-------------------------------------Yes 2431 1451 33 Check for Ascending MAT/MF/MT Order--------------------------Yes 2431 1451 34 Check for Legal MF/MT Numbers--------------------------------Yes 2431 1451 35 Allow Creation of Missing Sections---------------------------No 2431 1451 36 Allow Insertion of Energy Points-----------------------------No 2431 1451 37 Create Uniform Energy Grid-----------------------------------No 2431 1451 38 Delete Section if Cross Section =0 at All Energies-----------Yes 2431 1451 39 ***************** Program GROUPIE (VERSION 2002-1) **************2431 1451 40 Unshielded Group Averages Using 640 Groups 2431 1451 41 Weighting Spectrum: Flat (Constant) Spectrum 2431 1451 42 1 451 46 12431 1451 43 2 151 4 12431 1451 44 3 16 30 12431 1451 45 33 16 26 12431 1451 46 2431 1 099999 2431 0 0 0 2.40520E+4 5.14943E+1 0 0 1 02431 2151 1 2.40520E+4 1.00000E+0 0 0 1 02431 2151 2 1.00000E+0 2.00000E+7 0 0 0 02431 2151 3 0.0 5.44480E-1 0 0 0 02431 2151 4 2431 2 099999 2431 0 0 0 2.40520E+4 5.14943E+1 0 0 0 02431 3 16 1 -1.20394E+7-1.20394E+7 0 0 1 792431 3 16 2 79 1 2431 3 16 3 12200000.0 .000136222 12300000.0 .002913303 12400000.0 .0090350002431 3 16 4 12500000.0 .018505000 12600000.0 .029945000 12700000.0 .0421942862431 3 16 5 12800000.0 .054457131 12900000.0 .070114000 13000000.0 .0871593332431 3 16 6 13100000.0 .104204667 13200000.0 .121073583 13300000.0 .1368840002431 3 16 7 13400000.0 .152518000 13500000.0 .168152000 13600000.0 .1837860002431 3 16 8 13700000.0 .200229500 13800000.0 .221530000 13900000.0 .2436400002431 3 16 9 14000000.0 .265750000 14100000.0 .287860000 14200000.0 .3092252502431 3 16 10 14300000.0 .326122000 14400000.0 .342274000 14500000.0 .3584260002431 3 16 11 14600000.0 .374578000 14700000.0 .390333833 14800000.0 .4037126672431 3 16 12 14900000.0 .416695333 15000000.0 .429678000 15100000.0 .4426606672431 3 16 13 15200000.0 .455643333 15300000.0 .468626000 15400000.0 .4816086672431 3 16 14 15500000.0 .491881500 15600000.0 .499444500 15700000.0 .5070075002431 3 16 15 15800000.0 .514570500 15900000.0 .522133500 16000000.0 .5296965002431 3 16 16 16100000.0 .537259500 16200000.0 .544822500 16300000.0 .5523855002431 3 16 17 16400000.0 .559948500 16500000.0 .566084500 16600000.0 .5707935002431 3 16 18 16700000.0 .575502500 16800000.0 .580211500 16900000.0 .5849205002431 3 16 19 17000000.0 .589629500 17100000.0 .594338500 17200000.0 .5990475002431 3 16 20 17300000.0 .603756500 17400000.0 .608465500 17500000.0 .6119980002431 3 16 21 17600000.0 .614354000 17700000.0 .616710000 17800000.0 .6190660002431 3 16 22 17900000.0 .621422000 18000000.0 .623778000 18100000.0 .6261340002431 3 16 23 18200000.0 .628490000 18300000.0 .630846000 18400000.0 .6332020002431 3 16 24 18500000.0 .635069000 18600000.0 .636447000 18700000.0 .6378250002431 3 16 25 18800000.0 .639203000 18900000.0 .640581000 19000000.0 .6419590002431 3 16 26 19100000.0 .643337000 19200000.0 .644715000 19300000.0 .6460930002431 3 16 27 19400000.0 .647471000 19500000.0 .648849000 19600000.0 .6502270002431 3 16 28 19700000.0 .651605000 19800000.0 .652983000 19900000.0 .6543610002431 3 16 29 20000000.0 0.0 2431 3 16 30 2431 3 099999 2431 0 0 0 2.40520E+4 5.14943E+1 0 0 0 1243133 16 1 0.000000+0 0.000000+0 0 16 0 1243133 16 2 0.000000+0 0.000000+0 1 5 136 16243133 16 3 1.000000-5 1.220000+7 1.245000+7 1.255000+7 1.265000+7 1.275000+7243133 16 4 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7 1.600000+7243133 16 5 1.700000+7 1.800000+7 1.900000+7 2.000000+7 0.000000+0 0.000000+0243133 16 6 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16 7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16 8 0.000000+0 4.031000-2 3.015000-2 2.011000-2 4.441000-3 1.005000-2243133 16 9 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16 10 0.000000+0 0.000000+0 0.000000+0 2.256000-2 1.504000-2 3.322000-3243133 16 11 7.519000-3 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16 12 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.003000-2 2.216000-3243133 16 13 5.015000-3 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16 14 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.088000-3 1.107000-3243133 16 15 7.104000-4 1.385000-4 3.054000-4 1.269000-4 1.050000-3 6.131000-4243133 16 16 5.750000-4 5.345000-4 5.072000-4 2.506000-3 0.000000+0 0.000000+0243133 16 17 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0243133 16 18 0.000000+0 1.823000-3 2.416000-4 5.329000-4 2.213000-4 1.832000-3243133 16 19 1.070000-3 1.003000-3 9.325000-4 8.849000-4 4.904000-4 3.163000-4243133 16 20 3.318000-4 3.571000-4 2.085000-4 1.956000-4 1.818000-4 1.725000-4243133 16 21 7.342000-4 2.818000-4 7.875000-4 4.599000-4 4.313000-4 4.009000-4243133 16 22 3.805000-4 4.986000-4 3.270000-4 1.910000-4 1.791000-4 1.665000-4243133 16 23 1.580000-4 3.980000-3 1.581000-3 1.482000-3 1.378000-3 1.308000-3243133 16 24 1.358000-3 8.658000-4 8.048000-4 7.637000-4 1.194000-3 7.548000-4243133 16 25 7.163000-4 1.032000-3 6.658000-4 9.292000-4 243133 16 26 243133 099999 2431 0 0 0 0 0 0 0 2.50550E+4 5.44661E+1 0 0 34 102525 1451 1 0.0 0.0 0 0 0 62525 1451 2 1.00000E+0 2.00000E+7 0 5 10 20022525 1451 3 3.00000E+2 0.0 1 0 157 42525 1451 4 25-Mn- 55 JAERI,ORNL EVAL-MAR88 K.SHIBATA 2525 1451 5 DIST-Feb2004 2525 1451 6 ----IRDF-2002 MATERIAL 2525 2525 1451 7 -----INCIDENT NEUTRON DATA 2525 1451 8 ------ENDF-6 FORMAT 2525 1451 9 ******************************************************************2525 1451 10 25-MN- 55 JAERI,ORNL EVAL-MAR88 K.SHIBATA 2525 1451 11 DIST-SEP 1 REV2-SEP98 20010926 2525 1451 12 ----ENDF/B-VI MATERIAL 2525 REVISION 2 2525 1451 13 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2525 1451 14 DOSIMETRY 2525 1451 15 ******************************************************************2525 1451 16 2525 1451 17 **************************************************************** 2525 1451 18 2525 1451 19 ENDF/B-VI MOD 3 Revision, June 2000, S.C. Frankle, R.C. Reedy, 2525 1451 20 P.G. Young (LANL) 2525 1451 21 2525 1451 22 The secondary gamma-ray spectrum for radiative capture (MF 12, 2525 1451 23 MT 102) has been updated for new experimental data at incident 2525 1451 24 neutron energies up to 100 keV. 2525 1451 25 The previous (MOD 2) pure continuum at thermal neutron energy 2525 1451 26 is replaced by 321 discrete photons. 2525 1451 27 The Q-value for radiative capture was also updated in File 3. 2525 1451 28 Details of these changes are described in Frankel et al. [Fr01]. 2525 1451 29 2525 1451 30 REFERENCES 2525 1451 31 2525 1451 32 [Fr01] S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos 2525 1451 33 National Laboratory Report, LA-13812 (2001). 2525 1451 34 2525 1451 35 **************************************************************** 2525 1451 36 2525 1451 37 ENDF/B-VI MOD 2 Revision, October 1997, V. McLane (NNDC) 2525 1451 38 2525 1451 39 1. Corrected residual nucleus and AWR in File 6, MT=22,107. 2525 1451 40 2. Updated file 1 comments and and corrected references. 2525 1451 41 2525 1451 42 **************************************************************** 2525 1451 43 2525 1451 44 ENDF/B-VI MOD 1 Evaluation, March 1988, K. Shibata (ORNL,JAERI) 2525 1451 45 2525 1451 46 File 1 General Information ------------------------------------ 2525 1451 47 MT=451 Descriptive data and dictionary 2525 1451 48 2525 1451 49 File 2 Resonance Parameters ----------------------------------- 2525 1451 50 MT=151 Resolved resonance parameters for MLBW formula. 2525 1451 51 The parameters of the lowest four resonances were taken 2525 1451 52 from the work of Macklin [1]. Others were taken from the 2525 1451 53 compilation of mughabghab et al.[2] except that the 2525 1451 54 parameters of two negative resonances were adjusted so as 2525 1451 55 to fit to experimental thermal cross sections. 2525 1451 56 Resonance region : 1.0E-5 eV to 100 keV. 2525 1451 57 2525 1451 58 Scattering radius: 5.15 fm 2525 1451 59 2525 1451 60 Calculated 2200-m/s cross sections and res. integrals 2525 1451 61 2200-M/S RES. INTEG. 2525 1451 62 Elastic 2.167 b - 2525 1451 63 Capture 13.413 b 11.73 b 2525 1451 64 Total 15.579 b - 2525 1451 65 2525 1451 66 File 3 Neutron Cross Sections --------------------------------- 2525 1451 67 MT=1 Total 2525 1451 68 Below 100 keV : no background 2525 1451 69 Above 100 keV : based on the experimental data [3,4,5]. 2525 1451 70 MT=2 Elastic scattering 2525 1451 71 (total) - (nonelastic cross section). 2525 1451 72 2525 1451 73 MT=102 Radiative capture cross section 2525 1451 74 Below 100 keV: Resonance parameters given (no background). 2525 1451 75 Above 100 keV: Based on the experimental data [11-15] 2525 1451 76 (taken from JENDL-2 [20]). 2525 1451 77 2525 1451 78 2525 1451 79 File 33. Covariance data 2525 1451 80 MT=102 2525 1451 81 Estimated from experimental data and calculation. 2525 1451 82 2525 1451 83 ---------------------------------------------------------------- 2525 1451 84 REFERENCES 2525 1451 85 2525 1451 86 [1] R.L. Macklin, Nucl.Sci.Eng. 89, 362 (1985). 2525 1451 87 [2] S. F. Mughabghab, M. Divadeenam, and N.E. Holden, Neutron 2525 1451 88 Cross Sections, Vol. 1, Part A (Academic Press, 1981). 2525 1451 89 [3] S. Cierjacks, P. Forti, D. Kopsch et al., "High Resolution 2525 1451 90 Total Cross Sections for Na, Cl, K, V, Mn and Co between 0.5 2525 1451 91 and 30 MeV", report KFK-1000 (1968). 2525 1451 92 [4] W.F.E. Pineo, M. Divadeenam, E.G. Bilpuch et al., Ann.Phys. 2525 1451 93 84, 165 (1974). 2525 1451 94 [5] J.B. Garg, J. Rainwater, and W.W. Havens Jr., Nucl.Sci.Eng. 2525 1451 95 65, 76 (1978). 2525 1451 96 [6] C.Y. Fu, "A Consistent Nuclear Model for Compound and Pre- 2525 1451 97 compound Reactions with Conservation of Angular Momentum," 2525 1451 98 Oak Ridge National Laboratory report ORNL/TM-7042 (1980). 2525 1451 99 [7] C.Y. Fu, private communication (1985). 2525 1451 100 [8] Zhou Enchen, Huo Junde, Zhou Chunmei et al., Nucl.Data 2525 1451 101 Sheets 44, 463 (1985). 2525 1451 102 [9] F.G. Perey, Phys.Rev. 131, 745 (1963). 2525 1451 103 [10] J.R. Huizenga, and G.J. Igo, Nucl.Phys. 29, 462 (1962). 2525 1451 104 [11] J.B. Garg, R.L. Macklin, and J. Halperin, Phys.Rev. C18, 2525 1451 105 2079 (1978). 2525 1451 106 [12] A.G. Dovbenko, V.E. Kolesov, V.P. Koroleva et al., At.En. 26,2525 1451 107 67 (1969). 2525 1451 108 [13] H.O. Menlove, K.L. Coop, H.A. Grench et al. Phys.Rev. 163, 2525 1451 109 1299 (1967). 2525 1451 110 [14] O. Schwerer, M. Winkler-Rohatsch, H. Warhanek et al., Nucl. 2525 1451 111 Phys. A264, 105 (1976). 2525 1451 112 [15] M. Budnar, F. Cvelbar, E. Hodgson et al., "Prompt Gamma-ray 2525 1451 113 Spectra and Integarted Cross Sections for the Radiative 2525 1451 114 Capture of 14 MeV Neutrons for 28 Natural Targets ...," 2525 1451 115 report INDC(YUG)-6 (1979). 2525 1451 116 [16] L. Colli, I. Iori, S. Micheletti et al., Nuovo.cim. 21, 966 2525 1451 117 (1962). 2525 1451 118 [17] S. Sudar and Csikai, Nucl.Phys. A319, 157 (1979). 2525 1451 119 [18] M. Diksic, P. Strohal, and I. Slaus, J.Inorg.Nucl.Chem. 36, 2525 1451 120 477 (1974). 2525 1451 121 [19] C.H. Wu, R. Woelfle, and S.M. Qaim, Nucl.Phys. A329, 63 2525 1451 122 (1979). 2525 1451 123 [20] T. Nakagawa, "Summary of JENDL-2 General Purpose File," 2525 1451 124 report JAERI-M 84-103 (1984). 2525 1451 125 2525 1451 126 ******************************************************************2525 1451 127 2525 1451 128 2525 1451 129 2525 1451 130 2525 1451 131 ******************************************************************2525 1451 132 ************************ C O N T E N T S *********************** 2525 1451 133 ***************** Program LINEAR (VERSION 2002-1) ***************2525 1451 134 For All Data Greater than 1.0000E-10 barns in Absolute Value 2525 1451 135 Data Linearized to Within an Accuracy of .100000000 per-cent 2525 1451 136 ***************** Program RECENT (VERSION 2002-1) ***************2525 1451 137 for All Data Greater than 1.0000E-10 barns in Absolute Value 2525 1451 138 Data Linearized to within an Accuracy of .100000000 per-cent 2525 1451 139 ***************** Program SIGMA1 (VERSION 2002-1) ***************2525 1451 140 Data Doppler Broadened to 300.000000 Kelvin 2525 1451 141 for All Data Greater than 1.0000E-10 barns in Absolute Value 2525 1451 142 Data Linearized to Within an Accuracy pf .100000000 per-cent 2525 1451 143 ***************** Program FIXUP (Version 2002-1) ****************2525 1451 144 Corrected ZA/AWR in All Sections-----------------------------Yes 2525 1451 145 Corrected Thresholds-----------------------------------------Yes 2525 1451 146 Extended Cross Sections to 20 MeV----------------------------No 2525 1451 147 Allow Cross Section Deletion---------------------------------No 2525 1451 148 Allow Cross Section Reconstruction---------------------------No 2525 1451 149 Make All Cross Sections Non-Negative-------------------------Yes 2525 1451 150 Delete Energies Not in Ascending Order-----------------------Yes 2525 1451 151 Deleted Duplicate Points-------------------------------------Yes 2525 1451 152 Check for Ascending MAT/MF/MT Order--------------------------Yes 2525 1451 153 Check for Legal MF/MT Numbers--------------------------------Yes 2525 1451 154 Allow Creation of Missing Sections---------------------------No 2525 1451 155 Allow Insertion of Energy Points-----------------------------No 2525 1451 156 Create Uniform Energy Grid-----------------------------------No 2525 1451 157 Delete Section if Cross Section =0 at All Energies-----------Yes 2525 1451 158 ***************** Program GROUPIE (VERSION 2002-1) **************2525 1451 159 Unshielded Group Averages Using 640 Groups 2525 1451 160 Weighting Spectrum: Flat (Constant) Spectrum 2525 1451 161 1 451 165 32525 1451 162 2 151 4 12525 1451 163 3 102 217 32525 1451 164 33 102 17 12525 1451 165 2525 1 099999 2525 0 0 0 2.50550E+4 5.44661E+1 0 0 1 02525 2151 1 2.50550E+4 1.00000E+0 0 0 1 02525 2151 2 1.00000E-5 1.00000E+5 0 0 0 02525 2151 3 2.50000E+0 5.15000E-1 0 0 0 02525 2151 4 2525 2 099999 2525 0 0 0 2.50550E+4 5.44661E+1 0 0 0 02525 3102 1 7.27050E+6 7.27050E+6 0 0 1 6412525 3102 2 641 1 2525 3102 3 .000100000 210.836355 .000105000 205.919990 .000110000 201.2493862525 3102 4 .000115000 196.945550 .000120000 191.914305 .000127500 186.3322702525 3102 5 .000135000 181.232244 .000142500 176.545422 .000150000 171.4729812525 3102 6 .000160000 166.188822 .000170000 161.380891 .000180000 156.9675742525 3102 7 .000190000 152.861356 .000200000 149.112342 .000210000 145.5851502525 3102 8 .000220000 142.308952 .000230000 139.265033 .000240000 135.6921232525 3102 9 .000255000 131.765845 .000270000 128.736134 .000280000 125.3635062525 3102 10 .000300000 121.250491 .000320000 117.526305 .000340000 114.1155532525 3102 11 .000360000 110.974047 .000380000 108.106258 .000400000 105.1164352525 3102 12 .000425000 102.056964 .000450000 99.2616168 .000475000 96.69878032525 3102 13 .000500000 94.2886053 .000525000 92.0858615 .000550000 90.00503242525 3102 14 .000575000 88.0698021 .000600000 86.0862994 .000630000 84.04577582525 3102 15 .000660000 82.1786685 .000690000 80.3875951 .000720000 78.48183612525 3102 16 .000760000 76.4384216 .000800000 74.5411534 .000840000 72.80240492525 3102 17 .000880000 71.1498209 .000920000 69.6348527 .000960000 68.18401602525 3102 18 .001000000 66.6802433 .001050000 65.1103649 .001100000 63.63859712525 3102 19 .001150000 62.2824700 .001200000 60.6827442 .001275000 58.92456722525 3102 20 .001350000 57.3171279 .001425000 55.8169724 .001500000 54.22319982525 3102 21 .001600000 52.5567297 .001700000 51.0354372 .001800000 49.62844332525 3102 22 .001900000 48.3430548 .002000000 47.1512904 .002100000 46.03299822525 3102 23 .002200000 45.0088670 .002300000 44.0295564 .002400000 42.91218462525 3102 24 .002550000 41.6705962 .002700000 40.6994470 .002800000 39.64420442525 3102 25 .003000000 38.3451997 .003200000 37.1630951 .003400000 36.08207962525 3102 26 .003600000 35.0960903 .003800000 34.1857100 .004000000 33.23684052525 3102 27 .004250000 32.2734093 .004500000 31.3941221 .004750000 30.57240332525 3102 28 .005000000 29.8184172 .005250000 29.1195793 .005500000 28.45904112525 3102 29 .005750000 27.8537655 .006000000 27.2181353 .006300000 26.58283802525 3102 30 .006600000 25.9816432 .006900000 25.4240492 .007200000 24.81770612525 3102 31 .007600000 24.1692362 .008000000 23.5761125 .008400000 23.01947612525 3102 32 .008800000 22.5015420 .009200000 22.0202748 .009600000 21.56235132525 3102 33 .010000000 21.0882598 .010500000 20.5870570 .011000000 20.12861052525 3102 34 .011500000 19.6912520 .012000000 19.1899854 .012750000 18.63505402525 3102 35 .013500000 18.1210420 .014250000 17.6495482 .015000000 17.14682352525 3102 36 .016000001 16.6192227 .017000001 16.1355174 .017999999 15.69395772525 3102 37 .018999999 15.2880961 .020000000 14.9073261 .021000000 14.55883792525 3102 38 .022000000 14.2305260 .023000000 13.9246739 .024000000 13.56547322525 3102 39 .025500000 13.1784088 .027000001 12.8717830 .028000001 12.53936132525 3102 40 .029999999 12.1291901 .032000002 11.7547360 .034000002 11.41274262525 3102 41 .035999998 11.1005590 .037999999 10.8126558 .039999999 10.51231402525 3102 42 .042500000 10.2075815 .045000002 9.92927915 .047499999 9.670116892525 3102 43 .050000001 9.43185152 .052499998 9.21135883 .055000000 9.002585112525 3102 44 .057500001 8.81134205 .059999999 8.61018310 .063000001 8.408838552525 3102 45 .066000000 8.21903201 .068999998 8.04215889 .071999997 7.850354602525 3102 46 .075999998 7.64490573 .079999998 7.45726121 .083999999 7.280971782525 3102 47 .088000000 7.11715255 .092000000 6.96557614 .096000001 6.820888452525 3102 48 .100000001 6.67151668 .104999997 6.51337408 .109999999 6.368653332525 3102 49 .115000002 6.23031740 .119999997 6.07166588 .127499998 5.896271252525 3102 50 .135000005 5.73360932 .142499998 5.58432775 .150000006 5.425221582525 3102 51 .159999996 5.25797835 .170000002 5.10499237 .180000007 4.965220922525 3102 52 .189999998 4.83697678 .200000003 4.71666452 .209999993 4.606885102525 3102 53 .219999999 4.50328170 .230000004 4.40673826 .239999995 4.295008192525 3102 54 .254999995 4.16988289 .270000011 4.07410511 .280000001 3.967682092525 3102 55 .300000012 3.83731116 .319999993 3.71906586 .340000004 3.611487862525 3102 56 .360000014 3.51295061 .379999995 3.42086005 .400000006 3.327718272525 3102 57 .425000012 3.23316859 .449999988 3.14339320 .474999994 3.061576532525 3102 58 .500000000 2.98658914 .524999976 2.91563291 .550000012 2.850697872525 3102 59 .574999988 2.78904017 .600000024 2.72618353 .629999995 2.662508792525 3102 60 .660000026 2.60220368 .689999998 2.54701552 .720000029 2.485490582525 3102 61 .759999990 2.42159634 .800000012 2.36390611 .839999974 2.307430882525 3102 62 .879999995 2.25618204 .920000017 2.20711669 .959999979 2.162282282525 3102 63 1.00000000 2.11396879 1.04999995 2.06455202 1.10000002 2.018342272525 3102 64 1.14999998 1.97485369 1.20000005 1.92504843 1.27499998 1.869240432525 3102 65 1.35000002 1.81817448 1.42499995 1.77151193 1.50000000 1.720990942525 3102 66 1.60000002 1.66833877 1.70000005 1.62061813 1.79999995 1.577501822525 3102 67 1.89999998 1.53633298 2.00000000 1.49872866 2.09999990 1.463377202525 3102 68 2.20000005 1.43110672 2.29999995 1.40033212 2.40000010 1.365060862525 3102 69 2.54999995 1.32611312 2.70000005 1.29558299 2.79999995 1.262322122525 3102 70 3.00000000 1.22151624 3.20000005 1.18446758 3.40000010 1.150730102525 3102 71 3.59999990 1.12030727 3.79999995 1.09130384 4.00000000 1.062216532525 3102 72 4.25000000 1.03156309 4.50000000 1.00380165 4.75000000 .9784791932525 3102 73 5.00000000 .954717865 5.25000000 .932782336 5.50000000 .9124324242525 3102 74 5.75000000 .893112795 6.00000000 .873754670 6.30000019 .8536232792525 3102 75 6.59999990 .835215775 6.90000010 .817826751 7.19999981 .7993922782525 3102 76 7.59999990 .779064006 8.00000000 .761046505 8.39999962 .7434164102525 3102 77 8.80000019 .727236009 9.19999981 .712784460 9.60000038 .6987963352525 3102 78 10.0000000 .683807654 10.5000000 .668537936 11.0000000 .6543557092525 3102 79 11.5000000 .640924109 12.0000000 .625546703 12.7500000 .6090308242525 3102 80 13.5000000 .593231212 14.2500000 .579253798 15.0000000 .5637573762525 3102 81 16.0000000 .548133372 17.0000000 .533438811 18.0000000 .5201794832525 3102 82 19.0000000 .508113566 20.0000000 .496805619 21.0000000 .4864475312525 3102 83 22.0000000 .476931241 23.0000000 .468038827 24.0000000 .4575678872525 3102 84 25.5000000 .446585630 27.0000000 .437515619 28.0000000 .4280531342525 3102 85 30.0000000 .416384327 32.0000000 .406194345 34.0000000 .3965543532525 3102 86 36.0000000 .388085766 38.0000000 .380562488 40.0000000 .3726209172525 3102 87 42.5000000 .364787283 45.0000000 .357664387 47.5000000 .3513309852525 3102 88 50.0000000 .345529990 52.5000000 .340454878 55.0000000 .3358198712525 3102 89 57.5000000 .331458462 60.0000000 .327492717 63.0000000 .3233689752525 3102 90 66.0000000 .319662376 69.0000000 .316611514 72.0000000 .3136077742525 3102 91 76.0000000 .310483052 80.0000000 .308103202 84.0000000 .3062908522525 3102 92 88.0000000 .304877496 92.0000000 .304129079 96.0000000 .3035916572525 3102 93 100.000000 .303798850 105.000000 .304492151 110.000000 .3060217312525 3102 94 115.000000 .307938709 120.000000 .311490802 127.500000 .3170589832525 3102 95 135.000000 .324343252 142.500000 .333241886 150.000000 .3461758562525 3102 96 160.000000 .364380821 170.000000 .387408867 180.000000 .4159159992525 3102 97 190.000000 .451716142 200.000000 .496626962 210.000000 .5534975502525 3102 98 220.000000 .626778179 230.000000 .722826596 240.000000 .8927626382525 3102 99 255.000000 1.21455095 270.000000 1.66066610 280.000000 2.787732062525 3102 100 300.000000 7.64695078 320.000000 32.0716173 340.000000 21.79580152525 3102 101 360.000000 4.92641131 380.000000 1.93454159 400.000000 .9692847862525 3102 102 425.000000 .554036796 450.000000 .365099359 475.000000 .2638073652525 3102 103 500.000000 .203644625 525.000000 .165091912 550.000000 .1390815832525 3102 104 575.000000 .120836024 600.000000 .106629084 630.000000 .0956494812525 3102 105 660.000000 .088011582 690.000000 .082796624 720.000000 .0792082212525 3102 106 760.000000 .077781743 800.000000 .079440657 840.000000 .0849612782525 3102 107 880.000000 .096843366 920.000000 .121259750 960.000000 .1771357282525 3102 108 1000.00000 .405906404 1050.00000 6.82454289 1100.00000 5.669811202525 3102 109 1150.00000 .362922980 1200.00000 .135326257 1275.00000 .0814805222525 3102 110 1350.00000 .067185336 1425.00000 .063830493 1500.00000 .0678397682525 3102 111 1600.00000 2.00709610 1700.00000 .091643683 1800.00000 .1131801142525 3102 112 1900.00000 .155612894 2000.00000 .231560731 2100.00000 .3637798062525 3102 113 2200.00000 .532485764 2300.00000 .564672971 2400.00000 .3616478762525 3102 114 2550.00000 .173581391 2700.00000 .101383890 2800.00000 .0626053172525 3102 115 3000.00000 .037614764 3200.00000 .025780154 3400.00000 .0193484292525 3102 116 3600.00000 .015524039 3800.00000 .013933781 4000.00000 .0121013302525 3102 117 4250.00000 .017858518 4500.00000 .010019391 4750.00000 .4341479382525 3102 118 5000.00000 .011191569 5250.00000 .011343136 5500.00000 .0131607792525 3102 119 5750.00000 .016506710 6000.00000 .024240580 6300.00000 .1880253652525 3102 120 6600.00000 .118551401 6900.00000 .368728840 7200.00000 .1541118822525 3102 121 7600.00000 .047031546 8000.00000 .051335651 8400.00000 .2067577862525 3102 122 8800.00000 .221878480 9200.00000 .042097695 9600.00000 .0199195722525 3102 123 10000.0000 .008594289 10500.0000 .071149886 11000.0000 .0037810102525 3102 124 11500.0000 .022854740 12000.0000 .069200279 12750.0000 .0020981072525 3102 125 13500.0000 .001627971 14250.0000 .046109203 15000.0000 .0016402262525 3102 126 16000.0000 .029458063 17000.0000 .130854113 18000.0000 .0502274762525 3102 127 19000.0000 .009456944 20000.0000 .070705073 21000.0000 .0243172022525 3102 128 22000.0000 .038367457 23000.0000 .036942031 24000.0000 .0226151772525 3102 129 25500.0000 .069917782 27000.0000 .038371766 28000.0000 .0150495922525 3102 130 30000.0000 .021166131 32000.0000 .026460201 34000.0000 .0565505602525 3102 131 36000.0000 .023746549 38000.0000 .004205812 40000.0000 .0324913042525 3102 132 42500.0000 .020375006 45000.0000 .011843840 47500.0000 .0081271672525 3102 133 50000.0000 .011519836 52500.0000 .022264406 55000.0000 .0169532242525 3102 134 57500.0000 .020321384 60000.0000 .016414420 63000.0000 .0229886212525 3102 135 66000.0000 .018082623 69000.0000 .032203939 72000.0000 .0126870832525 3102 136 76000.0000 .008990626 80000.0000 .015362294 84000.0000 .0230386352525 3102 137 88000.0000 .013447072 92000.0000 .007114579 96000.0000 .0126574952525 3102 138 100000.000 .013806739 105000.000 .013550256 110000.000 .0136299922525 3102 139 115000.000 .013850078 120000.000 .014341383 127500.000 .0145210592525 3102 140 135000.000 .011994399 142500.000 .007991463 150000.000 .0078895692525 3102 141 160000.000 .008206275 170000.000 .008248789 180000.000 .0077906912525 3102 142 190000.000 .007096819 200000.000 .006509434 210000.000 .0061033842525 3102 143 220000.000 .006045272 230000.000 .005877501 240000.000 .0054732062525 3102 144 255000.000 .005270608 270000.000 .005672634 280000.000 .0062395712525 3102 145 300000.000 .005588775 320000.000 .005429058 340000.000 .0054481402525 3102 146 360000.000 .004134608 380000.000 .004576443 400000.000 .0054508182525 3102 147 425000.000 .004852925 450000.000 .005400591 475000.000 .0045650072525 3102 148 500000.000 .003995238 525000.000 .004095028 550000.000 .0044449782525 3102 149 575000.000 .004432601 600000.000 .003668962 630000.000 .0037604722525 3102 150 660000.000 .004102263 690000.000 .003633306 720000.000 .0033332162525 3102 151 760000.000 .003279851 800000.000 .003243539 840000.000 .0032055942525 3102 152 880000.000 .003123043 920000.000 .002977128 960000.000 .0027922012525 3102 153 1000000.00 .002620025 1100000.00 .002490003 1200000.00 .0023800022525 3102 154 1300000.00 .002284992 1400000.00 .002139233 1500000.00 .0019950132525 3102 155 1600000.00 .001940005 1700000.00 .001914998 1800000.00 .0018799982525 3102 156 1900000.00 .001835002 2000000.00 .001802005 2100000.00 .0017860032525 3102 157 2200000.00 .001770001 2300000.00 .001753999 2400000.00 .0017379982525 3102 158 2500000.00 .001721996 2600000.00 .001705994 2700000.00 .0016849272525 3102 159 2800000.00 .001633472 2900000.00 .001577164 3000000.00 .0015300022525 3102 160 3100000.00 .001490001 3200000.00 .001451001 3300000.00 .0014180012525 3102 161 3400000.00 .001386001 3500000.00 .001356001 3600000.00 .0013280012525 3102 162 3700000.00 .001300509 3800000.00 .001276061 3900000.00 .0012521212525 3102 163 4000000.00 .001228181 4100000.00 .001204868 4200000.00 .0011828282525 3102 164 4300000.00 .001160882 4400000.00 .001139808 4500000.00 .0011190302525 3102 165 4600000.00 .001098252 4700000.00 .001077733 4800000.00 .0010584212525 3102 166 4900000.00 .001039262 5000000.00 .001020102 5100000.00 .0010009422525 3102 167 5200000.00 .000982035 5300000.00 .000964328 5400000.00 .0009467772525 3102 168 5500000.00 .000930649 5600000.00 .000915945 5700000.00 .0009012422525 3102 169 5800000.00 .000886609 5900000.00 .000873163 6000000.00 .0008601392525 3102 170 6100000.00 .000847563 6200000.00 .000835951 6300000.00 .0008243712525 3102 171 6400000.00 .000812791 6500000.00 .000802101 6600000.00 .0007923012525 3102 172 6700000.00 .000782501 6800000.00 .000772701 6900000.00 .0007629012525 3102 173 7000000.00 .000754501 7100000.00 .000747501 7200000.00 .0007405012525 3102 174 7300000.00 .000733501 7400000.00 .000726501 7500000.00 .0007211012525 3102 175 7600000.00 .000717301 7700000.00 .000713501 7800000.00 .0007097012525 3102 176 7900000.00 .000705901 8000000.00 .000703201 8100000.00 .0007016012525 3102 177 8200000.00 .000700001 8300000.00 .000698400 8400000.00 .0006968002525 3102 178 8500000.00 .000695633 8600000.00 .000694900 8700000.00 .0006941672525 3102 179 8800000.00 .000693433 8900000.00 .000692700 9000000.00 .0006919672525 3102 180 9100000.00 .000691233 9200000.00 .000690500 9300000.00 .0006897672525 3102 181 9400000.00 .000689033 9500000.00 .000688300 9600000.00 .0006875672525 3102 182 9700000.00 .000686833 9800000.00 .000686100 9900000.00 .0006853672525 3102 183 10000000.0 .000684825 10100000.0 .000684475 10200000.0 .0006841252525 3102 184 10300000.0 .000683775 10400000.0 .000683425 10500000.0 .0006830752525 3102 185 10600000.0 .000682725 10700000.0 .000682375 10800000.0 .0006820252525 3102 186 10900000.0 .000681675 11000000.0 .000681325 11100000.0 .0006809752525 3102 187 11200000.0 .000680625 11300000.0 .000680275 11400000.0 .0006799252525 3102 188 11500000.0 .000679575 11600000.0 .000679225 11700000.0 .0006788752525 3102 189 11800000.0 .000678525 11900000.0 .000678175 12000000.0 .0006776852525 3102 190 12100000.0 .000677056 12200000.0 .000676426 12300000.0 .0006757972525 3102 191 12400000.0 .000675167 12500000.0 .000674537 12600000.0 .0006739082525 3102 192 12700000.0 .000673278 12800000.0 .000672649 12900000.0 .0006720192525 3102 193 13000000.0 .000671390 13100000.0 .000670760 13200000.0 .0006701302525 3102 194 13300000.0 .000669501 13400000.0 .000668871 13500000.0 .0006682422525 3102 195 13600000.0 .000667612 13700000.0 .000666925 13800000.0 .0006659192525 3102 196 13900000.0 .000664866 14000000.0 .000663812 14100000.0 .0006627592525 3102 197 14200000.0 .000661705 14300000.0 .000660651 14400000.0 .0006595982525 3102 198 14500000.0 .000658544 14600000.0 .000657491 14700000.0 .0006564372525 3102 199 14800000.0 .000655383 14900000.0 .000654330 15000000.0 .0006532762525 3102 200 15100000.0 .000652135 15200000.0 .000650786 15300000.0 .0006494292525 3102 201 15400000.0 .000648072 15500000.0 .000646714 15600000.0 .0006453572525 3102 202 15700000.0 .000643970 15800000.0 .000642400 15900000.0 .0006408002525 3102 203 16000000.0 .000638800 16100000.0 .000636400 16200000.0 .0006340002525 3102 204 16300000.0 .000631599 16400000.0 .000629199 16500000.0 .0006259992525 3102 205 16600000.0 .000621999 16700000.0 .000617999 16800000.0 .0006139982525 3102 206 16900000.0 .000609998 17000000.0 .000604498 17100000.0 .0005974982525 3102 207 17200000.0 .000590499 17300000.0 .000583499 17400000.0 .0005764992525 3102 208 17500000.0 .000568499 17600000.0 .000559499 17700000.0 .0005504992525 3102 209 17800000.0 .000541499 17900000.0 .000532499 18000000.0 .0005225602525 3102 210 18100000.0 .000511682 18200000.0 .000500805 18300000.0 .0004899272525 3102 211 18400000.0 .000479172 18500000.0 .000467322 18600000.0 .0004539702525 3102 212 18700000.0 .000440765 18800000.0 .000428381 18900000.0 .0004161262525 3102 213 19000000.0 .000401843 19100000.0 .000385857 19200000.0 .0003706312525 3102 214 19300000.0 .000355930 19400000.0 .000341972 19500000.0 .0003253282525 3102 215 19600000.0 .000306371 19700000.0 .000288758 19800000.0 .0002721502525 3102 216 19900000.0 .000256641 20000000.0 0.0 2525 3102 217 2525 3 099999 2525 0 0 0 2.50550E+4 5.44661E+1 0 0 0 1252533102 1 0.000000+0 0.000000+0 0 102 0 3252533102 2 0.000000+0 0.000000+0 0 1 8 4252533102 3 1.000000-5 1.600000-3 1.000000+3 2.500000-3 2.000000+6 1.000000-2252533102 4 2.000000+7 0.000000+0 252533102 5 0.000000+0 0.000000+0 0 1 26 13252533102 6 1.000000-5 3.960000-4 2.530000-2 2.227500-4 5.000000+1 3.960000-4252533102 7 2.000000+3 3.564000-3 6.000000+4 6.336000-3 1.500000+5 9.900000-3252533102 8 1.000000+6 2.227500-2 2.000000+6 2.227500-2 5.000000+6 3.960000-2252533102 9 7.000000+6 3.960000-2 1.000000+7 8.910000-2 1.450000+7 8.910000-2252533102 10 2.000000+7 0.000000+0 252533102 11 0.000000+0 0.000000+0 0 8 26 13252533102 12 1.000000-5 1.960000-2 2.530000-2 6.692000-3 5.000000+1 1.090900-6252533102 13 2.000000+3 2.851600-3 6.000000+4 1.747200-7 1.500000+5 5.975300-9252533102 14 1.000000+6 1.802000-9 2.000000+6 7.37120-10 5.000000+6 4.24360-10252533102 15 7.000000+6 2.29830-10 1.000000+7 4.22300-10 1.450000+7 3.92040-10252533102 16 2.000000+7 0.000000+0 252533102 17 252533 099999 2525 0 0 0 0 0 0 0 2.60540E+4 5.34762E+1 0 0 34 102625 1451 1 0.0 0.0 0 0 0 62625 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022625 1451 3 3.00000E+2 0.0 1 0 527 82625 1451 4 26-Fe- 54 FEI/LANL EVAL-Jun01 K.I.Zolotarev,M.Chadwick et al. 2625 1451 5 DIST-Feb2004 2625 1451 6 ----IRDF-2002 MATERIAL 2625 2625 1451 7 -----INCIDENT NEUTRON DATA 2625 1451 8 ------ENDF-6 FORMAT 2625 1451 9 ***************************************************************** 2625 1451 10 26-Fe- 54 FEI EVAL-Jun01 K.I.Zolotarev 2625 1451 11 DIST-May03 Mod1 20030512 2625 1451 12 ----BROND-2 MATERIAL 2625 REVISION 1 2625 1451 13 -----INCIDENT NEUTRON DATA 2625 1451 14 ------ENDF-6 FORMAT 2625 1451 15 ------Russian Reactor Dosimetry File RRDF-2002 2625 1451 16 ***************************************************************** 2625 1451 17 Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko 2625 1451 18 MF=3 2625 1451 19 MT= 16 - (n,2n) cross section 2625 1451 20 MT=107 - (n,a) cross section 2625 1451 21 ***************************************************************** 2625 1451 22 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2625 1451 23 DOSIMETRY 2625 1451 24 ******************************************************************2625 1451 25 26-Fe- 54 LANL,ORNL EVAL-SEP96 M.B.CHADWICK,P.G.YOUNG,D.HETRICK 2625 1451 26 Ch99,He91,Fu91 DIST-SEP 1 REV4- 20010926 2625 1451 27 ----ENDF/B-VI MATERIAL 2625 REVISION 4 2625 1451 28 MF=3 2625 1451 29 MT= 103 - (n,p) cross section 2625 1451 30 ******************************************************************2625 1451 31 ******** Start of (N,2N), (N,A) bibliographical component ******* 2625 1451 32 ***************************************************************** 2625 1451 33 ------Russian Reactor Dosimetry File RRDF-2002 2625 1451 34 ***************************************************************** 2625 1451 35 Authors of evaluation: K.I.Zolotarev and A.B.Pashchenko 2625 1451 36 MF=3 2625 1451 37 MT= 16 - (n,2n) cross section 2625 1451 38 ------------------------------------- 2625 1451 39 Excitation function for the Fe-54(n,2n)Fe-53m+g reaction in 2625 1451 40 the energy region from threshold to 21 MeV was evaluated by means 2625 1451 41 of statistical analysis of experimental cross section data [1-16] 2625 1451 42 and data from GNASH [31-32] calculation. 2625 1451 43 All experimental data if it was possible were renormalized to 2625 1451 44 the new standards for monitor reactions cross sections and decay 2625 1451 45 data. Decay data for the iron-53 were taken from ref. [17]. Recom-2625 1451 46 mended cross sections for the reaction Al-27(n,p)Mg-27 used as a 2625 1451 47 monitor in the measurements [7-8], [11-14] and [16] were taken 2625 1451 48 from new evaluation [18]. 2625 1451 49 Special correction was done with experimental data [1], [6]. 2625 1451 50 Data of Terrel and Holm [1] measured at 16.89 and 17.89 MeV 2625 1451 51 have been corrected introducing the Fc= 0.5 coefficient, because 2625 1451 52 of authors used double decreased value for positron yield. 2625 1451 53 Data of Andreev et al.[6] were renormalized using preliminary 2625 1451 54 evaluated integral of cross section value in the neutron energy 2625 1451 55 range from 15.0 to 16.5 MeV. Renormalised experimental data of 2625 1451 56 Andreev et al. are well consistent with Ryves measurements [9-10] 2625 1451 57 and with new experimental data for Fe-54(n,2n)Fe-53m+g reaction 2625 1451 58 obtained by Fessler [16]. 2625 1451 59 Experimental data from ref.[19-27] were rejected due to their 2625 1451 60 discrepancy with the main bulk of experimental data [1-16] and 2625 1451 61 data from theoretical model calculation. 2625 1451 62 The final procedure of evaluation Fe54(n,2n)Fe53m+g excitati- 2625 1451 63 on function from threshold to 21 MeV has been carried out within 2625 1451 64 the framework of generalized least squares method. Rational func- 2625 1451 65 tion was used as model function [28]. 2625 1451 66 U-235 thermal fission [29] and Cf-252 spontaneous fission 2625 1451 67 neutron spectra [30] averaged cross sections calculated from the 2625 1451 68 evaluated Fe54(n,2n)Fe53m+g excitation function are the following:2625 1451 69 2625 1451 70 ---------------------------------------------- 2625 1451 71 TYPE OF SPECTRUM I , mb (calc.) 2625 1451 72 ---------------------------I------------------ 2625 1451 73 U-235 neutron fission I 1.2839E-3 2625 1451 74 CF-252 spontan. fission I 3.6219E-3 2625 1451 75 ---------------------------------------------- 2625 1451 76 2625 1451 77 2625 1451 78 MT=107 - (n,a) cross section 2625 1451 79 ------------------------------------- 2625 1451 80 Excitation function for the Fe-54(n,a)Cr-51 reaction in the 2625 1451 81 energy region from 3 MeV to 20 MeV was evaluated by means of 2625 1451 82 statistical analysis of experimental cross section data [3], [4], 2625 1451 83 [11], [15], [21], [34-49] and data from GNASH [31-32] calculation.2625 1451 84 All experimental data were renormalized to the new standards 2625 1451 85 for monitor reactions cross sections and decay data. Experimental 2625 1451 86 data of S.R.Salisbury and R.A.Chalmers [5] were rejected due to 2625 1451 87 their inconsistency with theoretical model calculations data in 2625 1451 88 the energy region 3-5 MeV. Data of Y.M.Gledenov et al. [50] at 2625 1451 89 7 MeV were rejected due to their very big discrepancy with measu- 2625 1451 90 rements [39,46,49] and data from theoretical model calculation. 2625 1451 91 The final procedure of evaluation Fe-54(n,a)Cr-51 excitation 2625 1451 92 function from 3 MeV to 20 MeV have been carried out within the 2625 1451 93 framework of generalized least squares method. Rational function 2625 1451 94 was used as a model function [28]. 2625 1451 95 Integral experimental data for U-235 neutron fission spectrum 2625 1451 96 [51-53] was used for testing evaluated Fe-54(n,a)Cr-51 excitation 2625 1451 97 function. The results of testing are given in Table 1. 2625 1451 98 Data for U-235 thermal fission neutron spectrum and Cf-252 2625 1451 99 spontaneous fission neutron spectrum were taken from ref.[29] and 2625 1451 100 [30], respectively. 2625 1451 101 Table 1. 2625 1451 102 --------------------------------------------------------------- 2625 1451 103 TYPE OF SPECTRUM I ,MB (calc.) I , MB (measured) 2625 1451 104 ----------------------I-----------------I---------------------- 2625 1451 105 I I 2625 1451 106 U-235 neutron fission I 0.8459 I 0.790 +- 0.119 [51] 2625 1451 107 I I 0.850 +- 0.100 [52] 2625 1451 108 I I 0.891 +- 0.070 [53] 2625 1451 109 I I 0.850 +- 0.050 [* ] 2625 1451 110 I I 2625 1451 111 ----------------------I-----------------I---------------------- 2625 1451 112 I I 2625 1451 113 CF-252 spont. fission I 1.1114 I 2625 1451 114 I I 2625 1451 115 --------------------------------------------------------------- 2625 1451 116 [*] - Averaged cross section value obtained from ref. [51-53] 2625 1451 117 2625 1451 118 MF=33 2625 1451 119 MT= 16 - (n,2n) cross section cov. matrix 2625 1451 120 ------------------------------------------ 2625 1451 121 Uncertainties in the evaluated excitation function for the 2625 1451 122 reaction Fe-54(n,2n)Fe-53m+g are given in the form of relative co-2625 1451 123 variance matrix for the 12-neutron energy groups (LB=5). Covari- 2625 1451 124 ance matrix of uncertainties was calculated simultaneously with 2625 1451 125 recommended cross section data by means of PADE-2 code [33]. 2625 1451 126 Eigenvalues of the 6-th digits relative covariance matrix 2625 1451 127 given in the 33-file are the following: 2625 1451 128 2625 1451 129 3.65349E-06 4.84130E-06 6.69851E-06 1.07852E-05 2625 1451 130 2.27150E-05 7.28051E-05 4.72794E-04 5.38548E-03 2625 1451 131 1.13987E-02 1.32443E-02 3.40784E-02 1.48451E-01 2625 1451 132 2625 1451 133 MT=107 - (n,a) cross section cov. matrix 2625 1451 134 ------------------------------------------ 2625 1451 135 Uncertainties in the evaluated excitation function for the 2625 1451 136 reaction Fe-54(n,a)Cr-51 are given in the form of relative covari-2625 1451 137 ance matrix for the 17-neutron energy groups (LB=5). Covariance 2625 1451 138 matrix of uncertainties was calculated simultaneously with 2625 1451 139 recommended cross section data by means of PADE-2 code [33]. 2625 1451 140 Eigenvalues of the 6-th digits relative covariance matrix 2625 1451 141 given in the 33-file are the following: 2625 1451 142 2625 1451 143 1.85377E-09 2.23510E-09 2.38367E-09 2.64130E-09 2625 1451 144 3.38901E-09 3.88584E-09 5.57195E-09 9.37333E-09 2625 1451 145 2.56042E-08 1.44977E-06 8.05497E-04 1.66197E-03 2625 1451 146 2.03640E-03 2.27545E-03 4.21154E-03 1.42809E-02 2625 1451 147 2.58358E-02 2625 1451 148 2625 1451 149 References : 2625 1451 150 1. J.Terrell, D.M.Holm Physical Review, v.109, p.2031, 1958 2625 1451 151 2. M.J.Depraz et al. Journal de Physique, v.21, p.377, May 1960 2625 1451 152 3. D.M.Chittenden et al. Physical Review, v.122, p.860, 1961 2625 1451 153 4. W.G.Cross et al. Prog. Report EANDC(CAN)-16, p., January 1963 2625 1451 154 3. W.G.Cross et al. Prog. Report EANDC(CAN)-16, p.1, Jan. 1963 2625 1451 155 5. S.R.Salisbury, R.A.Chalmers Phys. Rev., v.B140, p.305, 1965 2625 1451 156 6. M.F.Andreev, V.I.Serov Jadernaja Fizika (Sov.), v.7, n.4, 2625 1451 157 p.745, April 1968 2625 1451 158 7. S.M.Qaim Nuclear Physics, v.A185, p.614, May 1972 2625 1451 159 8. R.A.Sigg, P.K.Kuroda Inorg. Nucl. Chem., v.37, p.631, 1975 2625 1451 160 9. T.B.Ryves et al. J. Metrologia, v.14, n.3, p.127, June 1978 2625 1451 161 10. T.B.Ryves et al. J. of Physics, pt.G, v.4, p.1783, 1978 2625 1451 162 11. B.M.Bahal, R.Pepelnik Report GKSS-84-E-, 1984 ; 2625 1451 163 B.M.Bahal, R.Pepelnik Progress Report NEANDC(E)-252/U,(5), 2625 1451 164 p.28, June 1984 2625 1451 165 12. L.R.Greenwood, R.K.Smither Proc. of Intern. Conf. on Nuclear 2625 1451 166 Data for Basic and Applied Science, Santa Fe, New Mexico, USA,2625 1451 167 13-17 May 1985, v.1, p.163 2625 1451 168 13. T.Katoh et al. Report JAERI-M-89-083, 1989 2625 1451 169 14. M.Viennot et al. Nucl. Sci. Eng., v.108, p.289, July 1991 2625 1451 170 15. A.Ercan et al. Proc. of International Conference on Nuclear 2625 1451 171 Data for Science and Technology, 13-17 May 1991, Julich, FRG, 2625 1451 172 Springer-Verlag, 1992, p.376-377 2625 1451 173 16. A.Fessler Report JUL-3502, FZ Julich GmbH, Germany, 1998 2625 1451 174 17. E.Browne, R.B.Firestone Table of Radioactive Isotopes, 2625 1451 175 John Wiley & Sons, New York, 1986 ; 2625 1451 176 R.B.Firestone Table of Isotopes, Eighth edition, Vol. 1, 2625 1451 177 John Wiley & Sons, Inc., New York, 1995 2625 1451 178 18. K.I.Zolotarev Report INDC(CCP)-431, Distr.: J+R/EL, IAEA, 2625 1451 179 Vienna, August 2002 2625 1451 180 19. D.L.Allan J. Proc. Phys. Soc., v.70, p.195, 1956 2625 1451 181 20. L.A.Rayburn Phys. Rev., v.122, p.168, 1961 2625 1451 182 21. H.Pollehn, H.Neuert Zeitschrift f. Naturforschung, sect.A, 2625 1451 183 v.16, p.227, 1961 2625 1451 184 22. C.Carles Comptes Rendus, v.257, p.659, July 1963 2625 1451 185 23. J.Csikai Report EANDC-50S, v.2, p.102, July 1965 2625 1451 186 24. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 2625 1451 187 25. J.Araminowicz, J.Dresler Rep.INR-1464, Swierk-Warsew,May 1973 2625 1451 188 26. M.Bormann et al. Zeitsch. f. Physik, sect.A, v.277, p.203, 2625 1451 189 June 1976 2625 1451 190 27. Zhou Muyao et al. Chinese J. of Nucl. Phys., v.9, p.34, 1987 2625 1451 191 28. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2625 1451 192 st's Meeting on Evaluation and Processing of Covariance Data, 2625 1451 193 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2625 1451 194 29. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2625 1451 195 Library, MAT-9228, MF=5, MT=18, eval. April 1989 2625 1451 196 30. W.Mannhart Report IAEA-TECDOC-410, p.158, 1987 2625 1451 197 31. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 2625 1451 198 Model Code for Calculation of Cross Section and Emission 2625 1451 199 Spectra. Report LA-6947, Los Alamos, 1977 2625 1451 200 32. E.L.Trykov, G.Ya.Tertychnyi Private communication, IPPE, 2625 1451 201 Obninsk, May 1999 2625 1451 202 33. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2625 1451 203 34. P.Venugopala Rao, R.W.Fink Phys. Rev., v.154, p.1023, 1967 2625 1451 204 35. S.M.Qaim et al. Conf. on Chemical Nuclear Data, Measurements 2625 1451 205 and Applicat., Univ. of Kent, Canterbury,20-22 September 1971,2625 1451 206 p.121 2625 1451 207 36. J.J.Singh Trans. Amer. Nucl. Soc., v.15, p.147, June 1972 2625 1451 208 37. G.N.Maslov et al. Sov. J. Yadernye Konstanty, v.9, p.50, 1972 2625 1451 209 38. K.Fukuda et al. Prog. Report NEANDC(J)-56/U, p.44, Sep. 1978 2625 1451 210 39. A.Paulsen et al. Nucl. Sci. Eng., v.72, v.1, p.113, 1979 2625 1451 211 40. O.I.Artem'ev et al. Atomnaja Energija (Sov.), v.49, no.3, 2625 1451 212 p.195, September 1980 2625 1451 213 41. L.R.Greenwood Progress Report DOE-ER-0046-21, p.15, May 1985 2625 1451 214 42. J.W.Meadows et al. Annals of Nuclear Energy, v.14, p.489, 2625 1451 215 September 1987 2625 1451 216 43. Y.Ikeda, C.Konno, K.Oishi et al. Report JAERI-1312, 1988 2625 1451 217 44. Lu Hanlin et al. Report INDC(CPR)-16, August 1989 2625 1451 218 45. S.K.Saraf et al. Nucl. Sci. Eng., v.107, p.365, April 1991 2625 1451 219 46. J.W.Meadows, W.Mannhart et.al. Proc. of Intern. Conference 2625 1451 220 on Nuclear Data for Science and Technology, 13-17 May 1991, 2625 1451 221 Julich, FRG, Springer-Verlag, 1992, p.288-290 2625 1451 222 47. A.Grallert et al. Report INDC(NDS)-286, p.193, IAEA, 1993 2625 1451 223 48. J.W.Meadows et al. Annals of Nuclear Energy, v.23, p.877, 2625 1451 224 July 1996 2625 1451 225 49. Lu Hanlin et al. Report INDC(CPR)-045, IAEA, Vienna, 1998 2625 1451 226 50. Y.M.Gledenov et al. Report INDC(CPR)-042, IAEA, Vienna, 1997 2625 1451 227 51. H.Braun, L.Nagy Radiochimica Acta, v.10, p.15, 1968 2625 1451 228 52. A.V.Bushuev et al. Atomnaja Energija (Sov.), v.63, no.3, 2625 1451 229 p.207, September 1987 2625 1451 230 53. O.Horibe et al. Proc. of Conference: 50 Years with Nuclear 2625 1451 231 Fission, Washington D.C., 25-28 April 1989, v.2, p.923 2625 1451 232 ******************************************************************2625 1451 233 ******** End of (N,2N), (N,A) bibliographical component ******** 2625 1451 234 ***************************************************************** 2625 1451 235 ******************************************************************2625 1451 236 ********* Start of (N,P) bibliographical component ********* 2625 1451 237 ***************************************************************** 2625 1451 238 2625 1451 239 ENDF/B-VI MOD 5 Revision, May 2000, S.C. Frankle, R.C. Reedy, 2625 1451 240 P.G. Young (LANL) 2625 1451 241 2625 1451 242 The secondary gamma-ray spectrum for radiative capture (MF 12, 2625 1451 243 MT 102) has been updated for new experimental data at incident 2625 1451 244 neutron energies up to 1 keV. 2625 1451 245 The MF=12, MT=102 yields above 1 keV were adjusted slightly to 2625 1451 246 force energy conservation. 2625 1451 247 The Q-value for radiative capture was also updated in File 3. 2625 1451 248 Details of these changes are described in Frankel et al. [Fr01]. 2625 1451 249 2625 1451 250 ******************************************************************2625 1451 251 2625 1451 252 ENDF/B-VI MOD 4 Evaluation, September 1997, M.B. Chadwick, 2625 1451 253 P.G. Young (LANL) and A.J. Koning (ECN) 2625 1451 254 2625 1451 255 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code 2625 1451 256 in cooperation with ECN Petten. 2625 1451 257 2625 1451 258 This evaluation provides a complete representation of the 2625 1451 259 nuclear data needed for transport, damage, heating, 2625 1451 260 radioactivity, and shielding applications over the incident 2625 1451 261 neutron energy range from 1.0E-11 to 150 MeV. The discussion 2625 1451 262 here is divided into the region below and above 20 MeV. 2625 1451 263 2625 1451 264 INCIDENT NEUTRON ENERGIES < 20 MeV 2625 1451 265 2625 1451 266 Below 20 MeV the evaluation is based completely on the ENDF/B- 2625 1451 267 VI.1 (Release 1) evaluation by [He91] (see also [Fu86]). The 2625 1451 268 following modifications were made to the ENDF/B-VI.1 evaluation: 2625 1451 269 2625 1451 270 1. The covariance files (MF=33) were removed from the file. 2625 1451 271 2625 1451 272 2. The derived MF=3 files for MT=203,205,207 were removed. 2625 1451 273 2625 1451 274 2625 1451 275 INCIDENT NEUTRON ENERGIES > 20 MeV 2625 1451 276 2625 1451 277 The evaluation above 20 MeV utilizes MF=6, MT=5 to represent 2625 1451 278 all reaction data. Production cross sections and emission 2625 1451 279 spectra are given for neutrons, protons, deuterons, tritons, 2625 1451 280 alpha particles, gamma rays, and all residual nuclides produced 2625 1451 281 (A>5) in the reaction chains. To summarize, the ENDF sections 2625 1451 282 with non-zero data above En = 20 MeV are: 2625 1451 283 2625 1451 284 MF=3 MT= 1 Total Cross Section 2625 1451 285 MT= 2 Elastic Scattering Cross Section 2625 1451 286 MT= 3 Nonelastic Cross Section 2625 1451 287 MT= 5 Sum of Binary (n,n') and (n,x) Reactions 2625 1451 288 2625 1451 289 MF=4 MT= 2 Elastic Angular Distributions 2625 1451 290 2625 1451 291 MF=6 MT= 5 Production Cross Sections and Energy-Angle 2625 1451 292 Distributions for Emission Neutrons, Protons, 2625 1451 293 Deuterons, and Alphas; and Angle-Integrated 2625 1451 294 Spectra for Gamma Rays and Residual Nuclei That 2625 1451 295 Are Stable Against Particle Emission 2625 1451 296 2625 1451 297 The evaluation is based on nuclear model calculations that 2625 1451 298 have been benchmarked to experimental data, especially for n + 2625 1451 299 Fe56 and p + Fe56 reactions [Ch96a]. We use the GNASH code 2625 1451 300 system [Yo92], which utilizes Hauser-Feshbach statistical, 2625 1451 301 preequilibrium and direct-reaction theories. Spherical optical 2625 1451 302 model calculations are used to obtain particle transmission 2625 1451 303 coefficients for the Hauser-Feshbach calculations, as well as 2625 1451 304 for the elastic neutron angular distributions. 2625 1451 305 Cross sections and spectra for producing individual residual 2625 1451 306 nuclei are included for reactions that exceed a cross section of 2625 1451 307 approximately 1 nb at any energy. The energy-angle-correlations 2625 1451 308 for all outgoing particles are based on Kalbach systematics 2625 1451 309 [Ka88]. 2625 1451 310 A model was developed to calculate the energy distributions of 2625 1451 311 all recoil nuclei in the GNASH calculations [Ch96b]. The recoil 2625 1451 312 energy distributions are represented in the laboratory system in 2625 1451 313 MT=5, MF=6, and are given as isotropic in the lab system. Note 2625 1451 314 that all other data in MT=5,MF=6 are given in the center-of-mass 2625 1451 315 system. This method of representation requires a modification of 2625 1451 316 the original ENDF-6 format. 2625 1451 317 Preequilibrium corrections were performed in the course of the 2625 1451 318 GNASH calculations using the exciton model of Kalbach [Ka77, 2625 1451 319 Ka85], validated by comparison with calculations using Feshbach, 2625 1451 320 Kerman, Koonin (FKK) theory [Ch93]. Discrete level data from 2625 1451 321 nuclear data sheets were matched to continuum level densities 2625 1451 322 using the formulation of Ignatyuk [Ig75] and pairing and shell 2625 1451 323 parameters from the Cook [Co67] analysis. Neutron and charged- 2625 1451 324 particle transmission coefficients were obtained from the 2625 1451 325 optical potentials, as discussed below. Gamma-ray transmission 2625 1451 326 coefficients were calculated using the Kopecky-Uhl model [Ko90]. 2625 1451 327 The neutron optical model potential of Arthur et al. [Ar80] 2625 1451 328 was used to calculate transmission coefficients and cross 2625 1451 329 sections with the SCAT2 code [Be92] up to a neutron energy of 26 2625 1451 330 MeV. Between 26 and 52 MeV, the imaginary volume component of 2625 1451 331 Arthur's potential was modified to better account for nonelastic 2625 1451 332 cross section measurements, and above 52 MeV the Semmering 2625 1451 333 potential of Madland [Ma88] was used. For protons, the 2625 1451 334 Beccetti-Greenlees potential [Be69] was utilized below 28 MeV, 2625 1451 335 and the Madland potential [Ma88] was used at higher energies. 2625 1451 336 The global spherical potential of Perey [Pe63] was utilized for 2625 1451 337 deuterons, and the potential of Beccetti-Greenlees [Be71] 2625 1451 338 was adopted for tritons. Finally, the alpha potential of Lemos 2625 1451 339 [Le72], as adapted by Arthur et al. [Ar80], was used for alpha 2625 1451 340 particles. 2625 1451 341 Direct reaction cross sections to discrete states were 2625 1451 342 calculated with the ECIS95 code [Re95] using deformation 2625 1451 343 parameters compiled in Nuclear Data Sheets. 2625 1451 344 We used the same values for the total, elastic, and nonelastic 2625 1451 345 cross sections (MF=3, MT=1,2,3) above 20 MeV as were used in the 2625 1451 346 n + 56Fe evaluation. 2625 1451 347 2625 1451 348 ***************************************************************** 2625 1451 349 2625 1451 350 REFERENCES 2625 1451 351 2625 1451 352 [Ar80] E.D. Arthur and P.G. Young, 'Evaluation of Neutron 2625 1451 353 Cross Sections to 40 MeV for 54,56Fe," Proc. Sym. on Neutron 2625 1451 354 Cross Sections from 10 to 50 MeV, 12-14 May 1980, Brookhaven 2625 1451 355 National Laboratory [Eds. M. R. Bhat and S. Pearlstein, BNL- 2625 1451 356 NCS-51245, 1980] p. 731. 2625 1451 357 [Be69] F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182 2625 1451 358 1190 (1969) 2625 1451 359 [Be71] F.D. Becchetti, Jr., and G.W. Greenlees, "Polarization 2625 1451 360 Phenomena in Nuclear Reactions," (Ed: H.H. Barschall and 2625 1451 361 W. Haeberli, The University of Wisconsin Press, 1971) p.682 2625 1451 362 [Be92] O. Bersillon, "SCAT2 - A Spherical Optical Model Code," 2625 1451 363 in Proc. ICTP Workshop on Computation and Analysis of Nuclear 2625 1451 364 Data Relevant to Nuclear Energy and Safety, 10 February-13 2625 1451 365 March, 1992, Trieste, Italy, to be published in World Scientific2625 1451 366 Press, and Progress Report of the Nuclear Physics Division, 2625 1451 367 Bruyeres-le-Chatel 1977, CEA-N-2037, p.111 (1978). 2625 1451 368 [Ch93] M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 2625 1451 369 [Ch96a] M.B. Chadwick and P.G. Young, "GNASH Calculations of 2625 1451 370 n,p + 54,56,57,58Fe and Benchmarking of Results" in APT 2625 1451 371 progress report: 1 August - 1 September 1996, internal Los 2625 1451 372 Alamos National Laboratory memo T-2-96/MS-52, 6 Aug. 1996 from 2625 1451 373 R.E. MacFarlane to L. Waters. 2625 1451 374 [Ch96b] M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J. 2625 1451 375 Koning, "High-Energy Nuclear Data Libraries for Accelerator- 2625 1451 376 Driven Technologies: Calculational Method for Heavy Recoils," 2625 1451 377 Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation 2625 1451 378 Technology and Applications, Kalmar, Sweden, 3-7 June 1996 2625 1451 379 [Ch99] M.B. Chadwick, P G. Young, G. M. Hale, et al., Los Alamos 2625 1451 380 National Laboratory report, LA-UR-99-1222 (1999) 2625 1451 381 [Co67] J.L. Cook, H. Ferguson, and A.R. deL Musgrove, Aust.J. 2625 1451 382 Phys. 20, 477 (1967) 2625 1451 383 [Fr01] S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos 2625 1451 384 National Laboratory Report, LA-13812 (2001). 2625 1451 385 [Fu86] C.Y. Fu and D.M. Hetrick, report ORNL/TM-9964 [ENDF-341] 2625 1451 386 (1986) 2625 1451 387 [He91] D.M. Hetrick, C.Y. Fu, and N.M. Larson, ENDF/B-VI.1 2625 1451 388 Evaluation of n + 54Fe, personal comm. (1991) 2625 1451 389 [Ig75] A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J. 2625 1451 390 Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485 2625 1451 391 (1975) 2625 1451 392 [Ka77] C. Kalbach, Z.Phys.A 283, 401 (1977) 2625 1451 393 [Ka85] C. Kalbach, Los Alamos report LA-10248-MS (1985) 2625 1451 394 [Ka88] C. Kalbach, Phys.Rev. C 37, 2350 (1988); see also 2625 1451 395 C. Kalbach and F.M. Mann, Phys.Rev. C 23, 112 (1981) 2625 1451 396 [Ko90] J. Kopecky and M. Uhl, Phys.Rev. C 41, 1941 (1990) 2625 1451 397 [Le72] O.F. Lemos, "Diffusion Elastique de Particules Alpha 2625 1451 398 de 21 a 29.6 MeV sur des Noyaux de la Region Ti-Zn," Orsay 2625 1451 399 report, Series A, No. 136 (1976). 2625 1451 400 [Ma88] D.G. Madland, Proc. OECD/NEANDC Specialist's Meeting on 2625 1451 401 Preequilibrium Nuclear Reactions, Semmering, Austria, Feb. 2625 1451 402 1988, NEANDC-245 (1988) p.103 2625 1451 403 [Pe63] C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963) 2625 1451 404 [Re95] J. Raynal, "Notes on ECIS94," CEA informal 2625 1451 405 report,Saclay (1995). 2625 1451 406 [Yo92] P.G. Young, E.D. Arthur, and M.B. Chadwick, Los Alamos 2625 1451 407 National Laboratory report LA-12343-MS (1992) 2625 1451 408 2625 1451 409 **************************************************************** 2625 1451 410 2625 1451 411 ENDF/B-VI MOD 3 Revision, October 1997, V. McLane (NNDC). 2625 1451 412 2625 1451 413 1. Corrected residual nucleus in File 6, MT=16,103. 2625 1451 414 2. Updated File 1 comments and corrected references. 2625 1451 415 2625 1451 416 **************************************************************** 2625 1451 417 2625 1451 418 ENDF/B-VI MOD 2 Revision, July 1991, (ORNL) 2625 1451 419 2625 1451 420 The elastic transformation matrix was removed. 2625 1451 421 2625 1451 422 **************************************************************** 2625 1451 423 2625 1451 424 ENDF/B-VI MOD 1 Evaluation, November 1989, D.M. Hetrick, C.Y. Fu,2625 1451 425 and N.M. Larson (ORNL) 2625 1451 426 2625 1451 427 This work employed nuclear model codes including the 2625 1451 428 Distorted Wave Born Approximation (DWBA) program DWUCK [1] 2625 1451 429 and the Hauser-Feshbach code TNG [2,3,4]. The TNG code provides 2625 1451 430 energy and angular distributions of particles emitted in the 2625 1451 431 compound and pre-compound reactions, ensures consistency among 2625 1451 432 all reactions, and maintains energy balance. details pertinent 2625 1451 433 to the contents of this evaluation will be published at a later 2625 1451 434 date. 2625 1451 435 Resonance parameters and thermal cross sections are from 2625 1451 436 Mughabghab [5] and adjusted through SAMMY by N. Larson using J.A.2625 1451 437 Harvey data [11] between 110 and 700 keV, capture cross section 2625 1451 438 background in File 3 was based on Allen [6]. Above 700 keV, 2625 1451 439 54Fe data from J.A. Harvey [11] used. 2625 1451 440 2625 1451 441 DESCRIPTION OF FILES 2625 1451 442 2625 1451 443 File 1 GENERAL INFORMATION ------------------------------------ 2625 1451 444 MT=451 General information, references, and definitions. 2625 1451 445 2625 1451 446 2625 1451 447 File 3 NEUTRON CROSS SECTIONS --------------------------------- 2625 1451 448 MT=103 (n,p) cross sections 2625 1451 449 Taken from the GLUCS [8] calculation in which this reaction 2625 1451 450 was studied simultaneously with 12 other dosimetry 2625 1451 451 reaction cross sections [9]. 2625 1451 452 2625 1451 453 File 33 UNCERTAINTY FILES ------------------------------------- 2625 1451 454 An LB=8 section is included for all non-derived files as 2625 1451 455 required by ENDF/B-VI. 2625 1451 456 MT=103 (n,p) covariances were taken from the GLUCS [8] 2625 1451 457 calculation in which this reaction was studied 2625 1451 458 simultaneously with 12 other dosimetry reaction cross 2625 1451 459 sections [9]. 2625 1451 460 ---------------------------------------------------------------- 2625 1451 461 REFERENCES: 2625 1451 462 2625 1451 463 [1] P.D. Kunz, "Distorted Wave Code DWUCK72," Univ. of Colorado, 2625 1451 464 unpublished (1972). 2625 1451 465 [2] C.Y. Fu, "A Consistent Nuclear Model For Compound and Pre- 2625 1451 466 compound Reactions with Conservation of Angular Momentum," 2625 1451 467 report ORNL/TM-7042 (1980), and Nucl.Sci.Eng. 100, 61 (1988).2625 1451 468 [3] C.Y Fu, "Development and Application of Multi-step 2625 1451 469 Hauser-Feshbach/Pre-equilibrium Model Theory," Symp. 2625 1451 470 Neutron Cross Sections from 10 to 50 MeV, Upton, N.Y., 2625 1451 471 May 12-14,1980, report BNL-NCS-51425, p.675. 2625 1451 472 [4] K. Shibata and C.Y. Fu, "Recent Improvements of the TNG 2625 1451 473 Statistical Model Code", report ORNL/TM-10093 (1986). 2625 1451 474 [5] S.F. Mughabghab, Neutron Cross Sections, Vol. 1: Resonance 2625 1451 475 Parameters and Thermal Cross Sections, Part A (Academic 2625 1451 476 Press, 1981). 2625 1451 477 [6] B.J. Allen, Geel Conference, 1977. 2625 1451 478 [7] P.T. Guenther, D.L. Smith, A.B. Smith et al., Ann.Nucl. En. 2625 1451 479 13 (11), 601-610 (1986). 2625 1451 480 [8] D.M. Hetrick and C.Y. Fu, "GLUCS: A Generalized Least Squares2625 1451 481 Program for Updating Cross Section Evaluations with 2625 1451 482 Correlated Data Sets," report ORNL/TM-7341 [ENDF-303] (1980).2625 1451 483 [9] C.Y. Fu and D.M. Hetrick, "Experience in Using the 2625 1451 484 Covariances of Some ENDF/B-V Dosimetry Cross Sections: 2625 1451 485 Proposed Improvements and Additions of Cross-reaction 2625 1451 486 Covariances," Proc. Fourth ASTM-Euratom Symp. on Reactor 2625 1451 487 Dosimetry, Gaithersburg, Maryland, March 22-26,1982, 2625 1451 488 (US National Bureau Of Standards, 1982) p.877. 2625 1451 489 ([0] S.M. Grimes, R.C. Haight, K.R. Alvar et al., Phys.Rev., C19, 2625 1451 490 2127 (1979). 2625 1451 491 [11] J.A. Harvey, private communication, 1989. 2625 1451 492 2625 1451 493 ******************************************************************2625 1451 494 ********* End of (N,P) bibliographical component ********* 2625 1451 495 ***************************************************************** 2625 1451 496 The Q values and threshold energies were updated prior to pro- 2625 1451 497 cessing through the codes to comply with the values obtained 2625 1451 498 using the NNDC calculation program which is based on the 1995 2625 1451 499 Update to the Atomic mass Evaluation. 2625 1451 500 2625 1451 501 File 2 added to the pointwise file containing only the effective 2625 1451 502 scattering radius with no resonance parameters given. 2625 1451 503 Taken from ENDF/B-VI 2625 1451 504 ************************ C O N T E N T S *********************** 2625 1451 505 2625 1451 506 ***************** Program LINEAR (VERSION 2002-1) ***************2625 1451 507 For All Data Greater than 1.0000E-10 barns in Absolute Value 2625 1451 508 Data Linearized to Within an Accuracy of .100000000 per-cent 2625 1451 509 ***************** Program SIGMA1 (VERSION 2002-1) ***************2625 1451 510 Data Doppler Broadened to 300.000000 Kelvin 2625 1451 511 for All Data Greater than 1.0000E-10 barns in Absolute Value 2625 1451 512 Data Linearized to Within an Accuracy pf .100000000 per-cent 2625 1451 513 ***************** Program FIXUP (Version 2002-1) ****************2625 1451 514 Corrected ZA/AWR in All Sections-----------------------------Yes 2625 1451 515 Corrected Thresholds-----------------------------------------Yes 2625 1451 516 Extended Cross Sections to 20 MeV----------------------------No 2625 1451 517 Allow Cross Section Deletion---------------------------------No 2625 1451 518 Allow Cross Section Reconstruction---------------------------No 2625 1451 519 Make All Cross Sections Non-Negative-------------------------Yes 2625 1451 520 Delete Energies Not in Ascending Order-----------------------Yes 2625 1451 521 Deleted Duplicate Points-------------------------------------Yes 2625 1451 522 Check for Ascending MAT/MF/MT Order--------------------------Yes 2625 1451 523 Check for Legal MF/MT Numbers--------------------------------Yes 2625 1451 524 Allow Creation of Missing Sections---------------------------No 2625 1451 525 Allow Insertion of Energy Points-----------------------------No 2625 1451 526 Create Uniform Energy Grid-----------------------------------No 2625 1451 527 Delete Section if Cross Section =0 at All Energies-----------Yes 2625 1451 528 ***************** Program GROUPIE (VERSION 2002-1) **************2625 1451 529 Unshielded Group Averages Using 640 Groups 2625 1451 530 Weighting Spectrum: Flat (Constant) Spectrum 2625 1451 531 1 451 539 12625 1451 532 2 151 4 12625 1451 533 3 16 25 12625 1451 534 3 103 70 12625 1451 535 3 107 62 12625 1451 536 33 16 21 12625 1451 537 33 103 57 12625 1451 538 33 107 35 12625 1451 539 2625 1 099999 2625 0 0 0 2.60540E+4 5.34760E+1 0 0 1 02625 2151 1 2.605400+4 1.000000+0 0 0 1 02625 2151 2 1.000000-5 7.000000+5 0 0 0 02625 2151 3 0.000000+0 5.480000-1 0 0 0 02625 2151 4 2625 2 099999 2625 0 0 0 2.60540E+4 5.34762E+1 0 0 0 02625 3 16 1 -1.33785E+7-1.33785E+7 0 0 1 652625 3 16 2 65 1 2625 3 16 3 13600000.0 .000103649 13700000.0 .000301866 13800000.0 .0005359992625 3 16 4 13900000.0 .000884865 14000000.0 .001354120 14100000.0 .0019488152625 3 16 5 14200000.0 .002673245 14300000.0 .003530825 14400000.0 .0045239552625 3 16 6 14500000.0 .005653915 14600000.0 .006920745 14700000.0 .0083231552625 3 16 7 14800000.0 .009858505 14900000.0 .011522700 15000000.0 .0133103002625 3 16 8 15100000.0 .015214450 15200000.0 .017227050 15300000.0 .0193389002625 3 16 9 15400000.0 .021539650 15500000.0 .023818050 15600000.0 .0261741002625 3 16 10 15700000.0 .028571900 15800000.0 .031006167 15900000.0 .0334769002625 3 16 11 16000000.0 .035947633 16100000.0 .038418367 16200000.0 .0408891002625 3 16 12 16300000.0 .043359833 16400000.0 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21 1.0000E-05 0.0000E+00 5.0000E+05 1.7503E-01 1.0000E+06 8.0475E-02262533103 22 2.0000E+06 3.2607E-02 4.0000E+06 2.7392E-02 6.0000E+06 2.5906E-02262533103 23 7.0000E+06 2.5300E-02 8.0000E+06 2.4784E-02 1.2000E+07 4.4955E-02262533103 24 1.4000E+07 3.3747E-02 1.6000E+07 9.9812E-02 2.0000E+07 0.0000E+00262533103 25 0.000000+0 0.000000+0 0 1 88 44262533103 26 1.000000-5 0.000000+0 6.900000+5 2.481500-3 8.000000+5 2.481500-3262533103 27 1.000000+6 5.828500-4 1.250000+6 5.828500-4 1.500000+6 5.828500-4262533103 28 1.750000+6 5.828500-4 2.000000+6 2.583600-4 2.250000+6 2.583600-4262533103 29 2.500000+6 2.583600-4 2.750000+6 2.583600-4 3.000000+6 2.583600-4262533103 30 3.220000+6 2.583600-4 3.250000+6 2.583600-4 3.350000+6 2.583600-4262533103 31 3.360000+6 2.583600-4 3.400000+6 2.583600-4 3.450000+6 2.583600-4262533103 32 3.500000+6 2.583600-4 3.750000+6 2.583600-4 4.000000+6 1.688300-4262533103 33 4.250000+6 1.688300-4 4.500000+6 1.688300-4 5.000000+6 1.688300-4262533103 34 5.500000+6 1.688300-4 6.000000+6 6.040000-4 7.000000+6 5.760900-4262533103 35 8.000000+6 1.713800-4 8.500000+6 1.713800-4 9.000000+6 1.713800-4262533103 36 9.500000+6 1.713800-4 1.000000+7 1.713800-4 1.050000+7 1.713800-4262533103 37 1.100000+7 1.713800-4 1.150000+7 1.713800-4 1.200000+7 5.274600-4262533103 38 1.250000+7 5.274600-4 1.300000+7 5.274600-4 1.350000+7 5.274600-4262533103 39 1.400000+7 3.382400-4 1.500000+7 3.382400-4 1.600000+7 9.862800-4262533103 40 1.800000+7 9.862800-4 2.000000+7 0.000000+0 262533103 41 0.000000+0 0.000000+0 0 8 88 44262533103 42 1.000000-5 0.000000+0 6.900000+5 2.35040-15 8.000000+5 1.71840-12262533103 43 1.000000+6 1.95530-11 1.250000+6 2.22450-10 1.500000+6 1.147100-9262533103 44 1.750000+6 1.375300-8 2.000000+6 2.922700-8 2.250000+6 7.618500-8262533103 45 2.500000+6 1.917900-7 2.750000+6 5.167200-7 3.000000+6 9.244100-7262533103 46 3.220000+6 1.009500-6 3.250000+6 1.023200-6 3.350000+6 1.076900-6262533103 47 3.360000+6 1.086200-6 3.400000+6 1.080800-6 3.450000+6 1.070700-6262533103 48 3.500000+6 1.331000-6 3.750000+6 2.066100-6 4.000000+6 1.708300-6262533103 49 4.250000+6 1.882200-6 4.500000+6 2.511700-6 5.000000+6 3.306500-6262533103 50 5.500000+6 3.915200-6 6.000000+6 1.542800-5 7.000000+6 1.456600-5262533103 51 8.000000+6 4.226500-6 8.500000+6 4.273700-6 9.000000+6 4.251200-6262533103 52 9.500000+6 4.188000-6 1.000000+7 4.185100-6 1.050000+7 4.171800-6262533103 53 1.100000+7 4.161300-6 1.150000+7 3.840400-6 1.200000+7 1.040900-5262533103 54 1.250000+7 9.461600-6 1.300000+7 8.428100-6 1.350000+7 7.345100-6262533103 55 1.400000+7 3.401000-6 1.500000+7 2.165000-6 1.600000+7 3.813700-6262533103 56 1.800000+7 2.035300-6 2.000000+7 0.000000+0 262533103 57 262533 099999 2.60540E+4 5.34762E+1 0 0 0 1262533107 1 0.000000+0 0.000000+0 0 107 0 1262533107 2 0.000000+0 0.000000+0 1 5 190 19262533107 3 1.000000-5 2.500000+6 5.000000+6 6.000000+6 7.000000+6 8.000000+6262533107 4 9.000000+6 1.000000+7 1.100000+7 1.200000+7 1.300000+7 1.400000+7262533107 5 1.450000+7 1.500000+7 1.600000+7 1.700000+7 1.800000+7 1.900000+7262533107 6 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533107 7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533107 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0262533107 9 0.000000+0 1.967980-2 9.633730-3 1.642170-3-7.873680-4 1.763730-4262533107 10 1.190170-3 1.391200-3 9.990920-4 4.107560-4-7.196720-6-7.635850-5262533107 11 2.317610-5 3.165150-4 6.879890-4 6.296170-4-2.098400-4-1.980280-3262533107 12 6.920690-3 2.102260-3-7.406730-5 2.004840-4 8.283510-4 1.023790-3262533107 13 8.128000-4 4.201790-4 9.153950-5-9.428760-6 2.000020-5 1.747130-4262533107 14 4.071900-4 4.020070-4-9.273340-5-1.202550-3 1.693190-3 1.036540-3262533107 15 6.190390-4 4.273150-4 3.244890-4 2.414200-4 1.586830-4 8.148000-5262533107 16 3.617670-5 1.695870-5 9.143770-6 3.129460-5 7.794820-5 1.265680-4262533107 17 1.582870-4 1.363610-3 1.012880-3 5.698220-4 2.544810-4 7.642630-5262533107 18 1.509480-6-8.578530-6 3.519450-6 1.342820-5 2.123660-5 1.402950-5262533107 19 -8.208480-6-2.787740-5-2.663740-5 1.050540-3 8.480750-4 5.713450-4262533107 20 3.081410-4 1.087930-4 6.133330-6-4.996150-6 1.701860-5 7.081030-5262533107 21 1.229820-4 8.175890-5-1.015020-4-4.373240-4 9.279640-4 8.104860-4262533107 22 5.821060-4 3.289850-4 1.294250-4 4.657870-5 3.281980-5 6.404150-5262533107 23 1.406320-4 1.699830-4 6.664270-5-2.138110-4 8.554680-4 7.401580-4262533107 24 5.261530-4 2.858140-4 1.331290-4 6.311590-5 2.314560-5 7.289320-5262533107 25 2.015140-4 3.382760-4 4.238150-4 7.539170-4 6.328030-4 4.126190-4262533107 26 2.242130-4 1.093590-4-4.162330-6-8.076930-6 1.879820-4 5.566300-4262533107 27 1.043390-3 6.135990-4 4.640330-4 2.936590-4 1.706290-4 2.345590-5262533107 28 -2.956160-5 1.618240-4 6.136620-4 1.290690-3 4.150570-4 3.211380-4262533107 29 2.403130-4 1.283090-4 6.183300-5 1.676890-4 4.809320-4 9.913180-4262533107 30 3.109670-4 2.900050-4 2.492490-4 2.077720-4 2.190780-4 3.096660-4262533107 31 4.856670-4 3.172600-4 3.412840-4 3.374690-4 2.897020-4 2.071610-4262533107 32 1.028700-4 4.633580-4 5.410820-4 4.564190-4 1.841470-4-2.618490-4262533107 33 7.451530-4 7.686590-4 5.507190-4 8.915960-5 1.151220-3 1.553840-3262533107 34 1.941310-3 3.226410-3 5.491560-3 1.062090-2 262533107 35 262533 099999 2625 0 0 0 0 0 0 0 2.60560E+4 5.54544E+1 0 0 34 102631 1451 1 0.0 0.0 0 0 0 62631 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022631 1451 3 3.00000E+2 0.0 1 0 261 42631 1451 4 26-Fe- 56 FEI EVAL-Oct02 K.I.Zolotarev 2631 1451 5 DIST-Feb2004 2631 1451 6 ----IRDF-2002 MATERIAL 2631 2631 1451 7 -----INCIDENT NEUTRON DATA 2631 1451 8 ------ENDF-6 FORMAT 2631 1451 9 ***************************************************************** 2631 1451 10 26-FE- 56 FEI EVAL-Oct02 K.I.Zolotarev 2631 1451 11 DIST-Nov03 20031105 2631 1451 12 ----BROND-3 MATERIAL 2631 Revision 2 2631 1451 13 -----INCIDENT NEUTRON DATA 2631 1451 14 ------ENDF-6 FORMAT 2631 1451 15 ------Russian Reactor Dosimetry File RRDF-2002 2631 1451 16 ***************************************************************** 2631 1451 17 Author of evaluation: K.I.Zolotarev 2631 1451 18 ***************************************************************** 2631 1451 19 MF=3 2631 1451 20 MT=103 -(n,p) cross section 2631 1451 21 ------------------------------------- 2631 1451 22 Excitation function for the Fe-56(n,p)Mn-56 reaction in the 2631 1451 23 energy region from threshold to 20 MeV was evaluated by means of 2631 1451 24 statistical analysis of experimental cross section data [1-40]. 2631 1451 25 The energy dependence of cross-section from 4.0 MeV to the 2631 1451 26 threshold was extrapolated with L=0 penetrability function for 2631 1451 27 the outgoing p + Mn-56 channel [41]. 2631 1451 28 Analised microscopic experimental data [1-40], [42-69] were 2631 1451 29 renormalized if it were possible to the new recommended standards 2631 1451 30 for monitor reaction cross sections and decay data. 2631 1451 31 Special correction was done with experimental data [2],[9-10] 2631 1451 32 and [17]. Experimental data of Terrell and Holm [2] corresponding 2631 1451 33 to neutron energies 6.54 MeV,7.41 MeV, 8.21 MeV were renormalized 2631 1451 34 to the preliminary evaluated cross section value at En=8.2 MeV. 2631 1451 35 Data of Liskien and Paulsen [9] measured in the energy range 2631 1451 36 12.60 - 19.58 MeV were corrected to the preliminary evaluated 2631 1451 37 integral of cross section in the energy interval 14-15 MeV. Data 2631 1451 38 obtained by Liskien and Paulsen in the energy range 6.06-8.20 MeV 2631 1451 39 [10] were renormalized to the preliminary evaluated cross section 2631 1451 40 value at 8.0 MeV. The correction factors for the experimental 2631 1451 41 data [2], [9] and [10] were Fc=1.23487, Fc=1.09674 and Fc=1.05835,2631 1451 42 respectively. Data of Smith and Meadows [17] obtained in the expe-2631 1451 43 riment with neutrons from D(d,n)He3 reaction in the energy range 2631 1451 44 6.486 - 9.945 MeV were renormalized to the preliminary evaluated 2631 1451 45 cross section value at 9.945 MeV. The correction factor for this 2631 1451 46 data was Fc=1.14066 . 2631 1451 47 Experimental information about Fe-56(n,p)Mn-56 reaction exci- 2631 1451 48 tation function are given in the ref. [13],[21],[23] in the form 2631 1451 49 of cross section ratios to monitor reactions. The results of pre- 2631 1451 50 cise relative measurements of Vonach et al. [13] in the energy 2631 1451 51 range 13.6-14.7 MeV were normalized to the preliminary evaluated 2631 1451 52 absolute cross section value 107.13 mb for Fe-56(n,p)Mn-56 reac- 2631 1451 53 tion at 14.7 MeV point. Raics et al. [21] and Antov et al. [23] 2631 1451 54 measured ratios of Fe-56(n,p)Mn-56 cross section to U-238(n,f) 2631 1451 55 and Al-27(n,a)Na-24 reactions cross section. Recommended absolute 2631 1451 56 cross section data for U-238(n,f) and Al-27(n,a)Na-24 reactions 2631 1451 57 were taken from ref. [70] and [71], respectively. 2631 1451 58 Experimental data from refs.[5] and [11] were used partially. 2631 1451 59 It were used only data obtained at 14.5 MeV [5] and in the energy 2631 1451 60 range 3.95 - 10.0 MeV [11]. Cross et al. data [5] for the neutron 2631 1451 61 energies 13.78 MeV, 14.07 MeV and 14.73 MeV were rejected due 2631 1451 62 to their systematically underestimation Fe-56(n,p)Mn-56 cross 2631 1451 63 section. The result of Grundl measurements at 14.1 MeV [11] was 2631 1451 64 not taken into account due to a very overestimated cross section 2631 1451 65 value obtained for this energy point. 2631 1451 66 Experimental cross section data [42-69] were rejected due to 2631 1451 67 their discrepancy with the main bulk of experimental data [1-40]. 2631 1451 68 In the rejected experiments [42-51], [53], [56-58], [60-62], 2631 1451 69 [64-67] and [69] the cross section values were measured only in a 2631 1451 70 one energy point in the interval 14 - 15 MeV. 2631 1451 71 Statistical analysis of input cross section data was carried 2631 1451 72 out by means of PADE-2 code [72]. Rational function was used as 2631 1451 73 the model function [73]. 2631 1451 74 U-235 thermal fission [74] and Cf-252 spontaneous fission 2631 1451 75 neutron spectra [75] averaged cross-sections calculated from the 2631 1451 76 evaluated Fe-56(n,p)Mn-56 excitation function are the following: 2631 1451 77 2631 1451 78 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2631 1451 79 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2631 1451 80 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2631 1451 81 U-235 neutron fission ³ 1.0997 ³ 1.09 +- 0.04 [76] 2631 1451 82 ³ ³ 1.13 +- 0.07 [77] 2631 1451 83 ³ ³ 1.083 +- 0.017 [78] 2631 1451 84 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2631 1451 85 Cf-252 spont. fission ³ 1.4626 ³ 1.471 +- 0.025 [79] 2631 1451 86 ³ ³ 1.465 +- 0.026 [80] 2631 1451 87 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2631 1451 88 2631 1451 89 MT=33 2631 1451 90 MT=103 -(n,p) cross section cov. matrix 2631 1451 91 --------------------------------------- 2631 1451 92 Uncertainties in the evaluated excitation function for the 2631 1451 93 reaction Fe-56(n,p)Mn-56 are given in the form of relative covari-2631 1451 94 ance matrix for the 25-neutron energy groups (LB=5). Covariance 2631 1451 95 matrix of uncertainties was calculated simultaneously with 2631 1451 96 recommended cross section data by means of PADE-2 code. 2631 1451 97 Eigenvalues of the 6-th digits relative covariance matrix 2631 1451 98 given in the 33-file are the following: 2631 1451 99 2631 1451 100 4.35195E-09 4.49367E-09 4.76919E-09 5.19445E-09 2631 1451 101 5.64595E-09 6.15408E-09 7.11594E-09 7.89561E-09 2631 1451 102 9.40785E-09 1.19394E-08 1.48739E-08 1.86574E-08 2631 1451 103 2.18905E-08 1.32640E-06 2.77587E-04 3.86559E-04 2631 1451 104 6.52483E-04 1.02062E-03 1.26078E-03 1.33119E-03 2631 1451 105 1.81407E-03 2.05264E-03 3.00443E-03 4.57086E-03 2631 1451 106 8.12954E-03 2631 1451 107 2631 1451 108 References : 2631 1451 109 1. G.Brown Philosophical Magazine, v.2, p.785, 1957 2631 1451 110 2. J.Terrell, D.M.Holm Phys. Rev., v.109, p.2031, 1958 2631 1451 111 3. H.Pollehn, H.Neuert Zeitschrift f. Naturforschung, sect. A, 2631 1451 112 v.16, p.227, 1961 2631 1451 113 4. 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Preprint FEI-1686, Obninsk, 1985 2631 1451 214 73. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2631 1451 215 st's Meeting on Evaluation and Processing of Covariance Data, 2631 1451 216 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2631 1451 217 74. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 2631 1451 218 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 2631 1451 219 75. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2631 1451 220 76. W.Mannhart Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry, 2631 1451 221 Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985 2631 1451 222 77. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2631 1451 223 Washington D.C., 25-28 April 1989, v.2, p.923 2631 1451 224 78. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 2631 1451 225 79. W.Mannhart Handbook on Nuclear Activation Cross Sections , 2631 1451 226 IAEA Tech. Report Ser. No.273, p.413, 1987 2631 1451 227 80. W.Mannhart Validation of Differential Cross Sections with 2631 1451 228 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2631 1451 229 September 2002 2631 1451 230 ***************************************************************** 2631 1451 231 The Q values and threshold energies were updated prior to pro- 2631 1451 232 cessing through the codes to comply with the values obtained 2631 1451 233 using the NNDC calculation program which is based on the 1995 2631 1451 234 Update to the Atomic mass Evaluation. 2631 1451 235 2631 1451 236 File 2 added to the pointwise file containing only the effective 2631 1451 237 scattering radius with no resonance parameters given. 2631 1451 238 Taken from ENDF/B-VI 2631 1451 239 ***************************************************************** 2631 1451 240 ***************** Program LINEAR (VERSION 2002-1) ***************2631 1451 241 For All Data Greater than 1.0000E-10 barns in Absolute Value 2631 1451 242 Data Linearized to Within an Accuracy of .100000000 per-cent 2631 1451 243 ***************** Program SIGMA1 (VERSION 2002-1) ***************2631 1451 244 Data Doppler Broadened to 300.000000 Kelvin 2631 1451 245 for All Data Greater than 1.0000E-10 barns in Absolute Value 2631 1451 246 Data Linearized to Within an Accuracy pf .100000000 per-cent 2631 1451 247 ***************** Program FIXUP (Version 2002-1) ****************2631 1451 248 Corrected ZA/AWR in All Sections-----------------------------Yes 2631 1451 249 Corrected Thresholds-----------------------------------------Yes 2631 1451 250 Extended Cross Sections to 20 MeV----------------------------No 2631 1451 251 Allow Cross Section Deletion---------------------------------No 2631 1451 252 Allow Cross Section Reconstruction---------------------------No 2631 1451 253 Make All Cross Sections Non-Negative-------------------------Yes 2631 1451 254 Delete Energies Not in Ascending Order-----------------------Yes 2631 1451 255 Deleted Duplicate Points-------------------------------------Yes 2631 1451 256 Check for Ascending MAT/MF/MT Order--------------------------Yes 2631 1451 257 Check for Legal MF/MT Numbers--------------------------------Yes 2631 1451 258 Allow Creation of Missing Sections---------------------------No 2631 1451 259 Allow Insertion of Energy Points-----------------------------No 2631 1451 260 Create Uniform Energy Grid-----------------------------------No 2631 1451 261 Delete Section if Cross Section =0 at All Energies-----------Yes 2631 1451 262 ***************** Program GROUPIE (VERSION 2002-1) **************2631 1451 263 Unshielded Group Averages Using 640 Groups 2631 1451 264 Weighting Spectrum: Flat (Constant) Spectrum 2631 1451 265 1 451 269 12631 1451 266 2 151 4 12631 1451 267 3 103 61 12631 1451 268 33 103 66 12631 1451 269 2631 1 099999 2631 0 0 0 2.60560E+4 5.54540E+1 0 0 1 02631 2151 1 2.605600+4 1.000000+0 0 0 1 02631 2151 2 1.000000-5 8.500000+5 0 0 0 02631 2151 3 0.000000+0 5.437300-1 0 0 0 02631 2151 4 2631 2 099999 2631 0 0 0 2.60560E+4 5.54544E+1 0 99 0 02631 3103 1 -2.91310E+6-2.91310E+6 0 0 1 1722631 3103 2 172 1 2631 3103 3 2900000.00 5.9516E-12 3000000.00 8.5021E-11 3100000.00 1.8579E-102631 3103 4 3200000.00 2.8657E-10 3300000.00 3.8734E-10 3400000.00 4.8811E-102631 3103 5 3500000.00 3.19099E-9 3600000.00 2.37217E-8 3700000.00 2.30222E-72631 3103 6 3800000.00 8.58242E-7 3900000.00 3.26945E-6 4000000.00 9.50803E-62631 3103 7 4100000.00 2.13588E-5 4200000.00 4.12059E-5 4300000.00 7.22242E-52631 3103 8 4400000.00 .000118637 4500000.00 .000186046 4600000.00 .0002818582631 3103 9 4700000.00 .000415817 4800000.00 .000600658 4900000.00 .0008528442631 3103 10 5000000.00 .001193315 5100000.00 .001648055 5200000.00 .0022481552631 3103 11 5300000.00 .003028750 5400000.00 .004026080 5500000.00 .0052718752631 3103 12 5600000.00 .006784800 5700000.00 .008560220 5800000.00 .0105616952631 3103 13 5900000.00 .012519200 6000000.00 .014238450 6100000.00 .0161200002631 3103 14 6200000.00 .018181150 6300000.00 .020072250 6400000.00 .0217811502631 3103 15 6500000.00 .023826650 6600000.00 .025746950 6700000.00 .0270865002631 3103 16 6800000.00 .028396700 6900000.00 .029706900 7000000.00 .0310583252631 3103 17 7100000.00 .032450975 7200000.00 .033907625 7300000.00 .0354282752631 3103 18 7400000.00 .037005975 7500000.00 .038640725 7600000.00 .0403027402631 3103 19 7700000.00 .041992020 7800000.00 .043681300 7900000.00 .0453705802631 3103 20 8000000.00 .047059860 8100000.00 .048700150 8200000.00 .0502914502631 3103 21 8300000.00 .051829775 8400000.00 .053315125 8500000.00 .0547361002631 3103 22 8600000.00 .056092700 8700000.00 .057449300 8800000.00 .0587423902631 3103 23 8900000.00 .059971970 9000000.00 .061201550 9100000.00 .0624311302631 3103 24 9200000.00 .063660710 9300000.00 .064888420 9400000.00 .0661142602631 3103 25 9500000.00 .067340100 9600000.00 .068565940 9700000.00 .0697917802631 3103 26 9800000.00 .071069213 9900000.00 .072398238 10000000.0 .0737272632631 3103 27 10100000.0 .075056288 10200000.0 .076448125 10300000.0 .0779027752631 3103 28 10400000.0 .079357425 10500000.0 .080812075 10600000.0 .0823283602631 3103 29 10700000.0 .083906280 10800000.0 .085484200 10900000.0 .0870621202631 3103 30 11000000.0 .088640040 11100000.0 .090238714 11200000.0 .0918581432631 3103 31 11300000.0 .093477571 11400000.0 .095097000 11500000.0 .0967164292631 3103 32 11600000.0 .098335857 11700000.0 .099955286 11800000.0 .1014997502631 3103 33 11900000.0 .102969250 12000000.0 .104438750 12100000.0 .1059082502631 3103 34 12200000.0 .107264833 12300000.0 .108508500 12400000.0 .1097521672631 3103 35 12500000.0 .110859000 12600000.0 .111829000 12700000.0 .1127990002631 3103 36 12800000.0 .113632500 12900000.0 .114329500 13000000.0 .1149057502631 3103 37 13100000.0 .115361250 13200000.0 .115689750 13300000.0 .1158912502631 3103 38 13400000.0 .115996000 13500000.0 .115908000 13600000.0 .1157240002631 3103 39 13700000.0 .115415750 13800000.0 .114983250 13900000.0 .1144076672631 3103 40 14000000.0 .113689000 14100000.0 .112970333 14200000.0 .1121030002631 3103 41 14300000.0 .111087000 14400000.0 .110071000 14500000.0 .1089375002631 3103 42 14600000.0 .107686500 14700000.0 .106435500 14800000.0 .1050908752631 3103 43 14900000.0 .103652625 15000000.0 .102214375 15100000.0 .1007761252631 3103 44 15200000.0 .099278768 15300000.0 .097722305 15400000.0 .0961658412631 3103 45 15500000.0 .094609377 15600000.0 .093052914 15700000.0 .0914964502631 3103 46 15800000.0 .089939986 15900000.0 .088383523 16000000.0 .0868270592631 3103 47 16100000.0 .085270595 16200000.0 .083714132 16300000.0 .0822083502631 3103 48 16400000.0 .080753250 16500000.0 .079298150 16600000.0 .0778430502631 3103 49 16700000.0 .076387950 16800000.0 .074994575 16900000.0 .0736629252631 3103 50 17000000.0 .072331275 17100000.0 .070999625 17200000.0 .0697291002631 3103 51 17300000.0 .068519700 17400000.0 .067310300 17500000.0 .0661009002631 3103 52 17600000.0 .064953475 17700000.0 .063868025 17800000.0 .0627825752631 3103 53 17900000.0 .061697125 18000000.0 .060671725 18100000.0 .0597063752631 3103 54 18200000.0 .058741025 18300000.0 .057775675 18400000.0 .0568670002631 3103 55 18500000.0 .056015000 18600000.0 .055163000 18700000.0 .0543110002631 3103 56 18800000.0 .053511638 18900000.0 .052764913 19000000.0 .0520181882631 3103 57 19100000.0 .051271463 19200000.0 .050573338 19300000.0 .0499238132631 3103 58 19400000.0 .049274288 19500000.0 .048624763 19600000.0 .0480200382631 3103 59 19700000.0 .047460113 19800000.0 .046900188 19900000.0 .0463402632631 3103 60 20000000.0 0.0 2631 3103 61 2631 3 099999 2631 0 0 0 2.60560E+4 5.54544E+1 0 0 0 1263133103 1 0.000000+0 0.000000+0 0 103 0 1263133103 2 0.000000+0 0.000000+0 1 5 378 27263133103 3 1.000000-5 2.900000+6 5.000000+6 6.000000+6 7.000000+6 8.000000+6263133103 4 9.000000+6 1.000000+7 1.100000+7 1.150000+7 1.200000+7 1.250000+7263133103 5 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7 1.550000+7263133103 6 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7 1.850000+7263133103 7 1.900000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0263133103 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 9 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 10 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0263133103 11 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 6.980350-3263133103 12 2.338270-3 1.702360-4 4.353220-4-6.002120-5 3.473780-4 1.342800-4263133103 13 -1.013590-4-1.453170-4-1.032110-4-1.910370-5 5.580590-5 8.566180-5263133103 14 6.229070-5 2.961970-6-6.242280-5-1.056550-4-1.094340-4-7.037860-5263133103 15 1.788090-6 8.621690-5 1.524240-4 1.605400-4 6.014590-5-2.124840-4263133103 16 2.467980-3 6.043630-4 3.578130-4 2.055220-4 1.680110-4 2.128390-4263133103 17 1.769540-4 1.366700-4 9.916600-5 7.004770-5 5.015720-5 3.742600-5263133103 18 2.898990-5 2.276790-5 1.801980-5 1.502850-5 1.442320-5 1.658450-5263133103 19 2.127870-5 2.748400-5 3.331180-5 3.591800-5 3.136190-5 1.435050-5263133103 20 2.089910-3 5.811850-4 5.051160-4 4.463010-4 2.985330-4 2.415270-4263133103 21 2.109770-4 1.764960-4 1.378030-4 9.933400-5 6.703660-5 4.523120-5263133103 22 3.491560-5 3.399020-5 3.868840-5 4.511620-5 5.024920-5 5.232580-5263133103 23 5.086090-5 4.651010-5 4.095130-5 3.684360-5 3.791020-5 2.324770-3263133103 24 8.909680-4 2.243430-4 3.463970-4 3.586080-4 2.979140-4 2.137770-4263133103 25 1.316880-4 7.020240-5 3.717340-5 2.982110-5 3.851070-5 5.200710-5263133103 26 6.154930-5 6.257380-5 5.457370-5 4.014340-5 2.400670-5 1.237870-5263133103 27 1.271590-5 3.381980-5 8.624440-5 1.319090-3 6.075680-4 2.355910-4263133103 28 1.831400-4 1.797820-4 1.702720-4 1.477000-4 1.151050-4 8.057420-5263133103 29 5.231750-5 3.522030-5 2.987090-5 3.362030-5 4.237480-5 5.207540-5263133103 30 5.950970-5 6.257730-5 6.025220-5 5.245210-5 3.991740-5 2.416230-5263133103 31 9.567740-4 6.313680-4 3.548430-4 2.308720-4 1.569670-4 1.197830-4263133103 32 1.026490-4 9.147350-5 7.836800-5 6.207430-5 4.573040-5 3.376500-5263133103 33 2.952730-5 3.417020-5 4.646570-5 6.302500-5 7.854620-5 8.588480-5263133103 34 7.587850-5 3.687780-5 8.246530-4 7.196220-4 5.696170-4 4.048130-4263133103 35 2.567500-4 1.473680-4 8.506790-5 6.460670-5 7.167200-5 8.965720-5263133103 36 1.051550-4 1.104520-4 1.034970-4 8.671160-5 6.568700-5 4.830190-5263133103 37 4.434860-5 6.563950-5 1.265420-4 7.702400-4 6.761850-4 5.244410-4263133103 38 3.568230-4 2.113590-4 1.126140-4 6.666460-5 6.340340-5 8.424730-5263133103 39 1.104340-4 1.283210-4 1.311310-4 1.184390-4 9.489290-5 6.907830-5263133103 40 5.291210-5 6.162700-5 1.143590-4 6.324280-4 5.237220-4 3.833900-4263133103 41 2.465290-4 1.404880-4 7.815570-5 5.731600-5 6.563580-5 8.760000-5263133103 42 1.098400-4 1.236410-4 1.252570-4 1.152190-4 9.746320-5 7.872510-5263133103 43 6.832810-5 7.841040-5 4.660800-4 3.703690-4 2.615780-4 1.637400-4263133103 44 9.329270-5 5.585710-5 4.758000-5 5.925580-5 8.050700-5 1.024550-4263133103 45 1.187560-4 1.256000-4 1.212720-4 1.056640-4 7.989500-5 4.610700-5263133103 46 3.221980-4 2.512470-4 1.743220-4 1.074110-4 6.116180-5 3.911000-5263133103 47 3.878810-5 5.417430-5 7.796430-5 1.029630-4 1.226310-4 1.310680-4263133103 48 1.227080-4 9.188330-5 3.231030-5 2.173730-4 1.684250-4 1.165000-4263133103 49 7.298230-5 4.527820-5 3.576720-5 4.260230-5 6.120560-5 8.556410-5263133103 50 1.089960-4 1.244090-4 1.241890-4 9.983000-5 4.136060-5 1.477430-4263133103 51 1.185060-4 8.879480-5 6.555160-5 5.284920-5 5.160670-5 6.020940-5263133103 52 7.533050-5 9.255920-5 1.067160-4 1.118820-4 1.012050-4 6.649160-5263133103 53 1.138400-4 1.048040-4 9.446080-5 8.548980-5 7.961900-5 7.752210-5263133103 54 7.899010-5 8.316230-5 8.870200-5 9.388370-5 9.659180-5 9.423730-5263133103 55 1.173920-4 1.242110-4 1.246130-4 1.193480-4 1.101060-4 9.908370-5263133103 56 8.873970-5 8.171760-5 8.089820-5 8.954360-5 1.115220-4 1.477160-4263133103 57 1.605150-4 1.612980-4 1.513370-4 1.337330-4 1.127760-4 9.360740-5263133103 58 8.216990-5 8.538600-5 1.115530-4 1.858920-4 1.970250-4 1.936170-4263133103 59 1.779120-4 1.539130-4 1.268590-4 1.030300-4 8.983880-5 9.619200-5263133103 60 2.208930-4 2.302800-4 2.251120-4 2.070470-4 1.789350-4 1.444760-4263133103 61 1.081310-4 7.522490-5 2.563740-4 2.685600-4 2.647750-4 2.436270-4263133103 62 2.040130-4 1.447640-4 6.437880-5 3.017630-4 3.184520-4 3.123860-4263133103 63 2.767690-4 2.037260-4 8.360900-5 3.586650-4 3.751200-4 3.563580-4263133103 64 2.888990-4 1.563190-4 4.204370-4 4.352960-4 4.045150-4 3.095700-4263133103 65 5.050460-4 5.556140-4 5.751470-4 7.485460-4 9.914130-4 1.606050-3263133103 66 263133 099999 2631 0 0 0 0 0 0 0 2.60580E+4 5.74356E+1 0 0 34 102637 1451 1 0.0 0.0 0 0 0 62637 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022637 1451 3 3.00000E+2 0.0 1 0 162 42637 1451 4 26-Fe- 58 JAERI EVAL-OCT95 T.NAKAGAWA 2637 1451 5 DIST-Feb2004 2637 1451 6 ----IRDF-2002 MATERIAL 2637 2637 1451 7 -----INCIDENT NEUTRON DATA 2637 1451 8 ------ENDF-6 FORMAT 2637 1451 9 ******************************************************************2637 1451 10 26-FE- 58 JAERI EVAL-OCT95 T.NAKAGAWA 2637 1451 11 DIST-JUL98 2637 1451 12 ----JENDL/D-99 MATERIAL 2637 2637 1451 13 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2637 1451 14 DOSIMETRY 2637 1451 15 ******************************************************************2637 1451 16 HISTORY 2637 1451 17 93-10 JENDL-3.2. 2637 1451 18 RE-EVALUATION WAS MADE BY 2637 1451 19 T.NAKAGAWA (NDC/JAERI): RESONANCE PARAMS, CROSS SECTIONS 2637 1451 20 S.IGARASI (NEDAC): GAMMA-RAY PRODUCTION DATA BELOW 10 KEV2637 1451 21 COMPILED BY T.NAKAGAWA 2637 1451 22 95-10 COVARIANCES WERE ESTIMATED BY T.NAKAGAWA. 2637 1451 23 98-04 COMPILED TO JENDL DOSIMETRY FILE 99. 2637 1451 24 02-10 COVARIANCES WERE MODIFIED BY K.SHIBATA. 2637 1451 25 2637 1451 26 ===== ORIGINAL DATA FILE ===== 2637 1451 27 2637 1451 28 2637 1451 29 FE-58 CAPTURE FE-59 (HALF-LIFE = 44.503 H) 2637 1451 30 2637 1451 31 MF=1 GENERAL INFORMATION 2637 1451 32 MT=451 DESCRIPTIVE DATA AND DICTIONARY 2637 1451 33 2637 1451 34 2637 1451 35 MF=2 RESONANCE PARAMETERS 2637 1451 36 MT = 151 RESOLVED RESONANCES 2637 1451 37 TAKEM FROM JENDL-3.2/1/. 2637 1451 38 RESONANCE REGION = 1.0E-5 EV TO 350.0 KEV 2637 1451 39 THE MULTILEVEL BREIT-WIGNER FORMULA WAS USED. PARAMETERS 2637 1451 40 WERE DETERMINED ON THE BASIS OF DATA BY GARG ET AL./2/, 2637 1451 41 KAEPPELER ET AL./3/, ALLEN AND MACKLIN/4/. 2637 1451 42 2637 1451 43 CALCULATED 2200-M/S CROSS SECTIONS AND RES. INTEGRALS. 2637 1451 44 2200-M/S RES. INTEG. 2637 1451 45 ELASTIC 6.470 B - 2637 1451 46 CAPTURE 1.300 B 1.36 B 2637 1451 47 TOTAL 7.770 B - 2637 1451 48 2637 1451 49 MF = 3 CROSS SECTIONS 2637 1451 50 MT = 1 TOTAL CROSS SECTION 2637 1451 51 TAKEN FROM JENDL-3.2/1/. 2637 1451 52 2637 1451 53 MT=2 ELASTIC 2637 1451 54 TOTAL CROSS SECTION - SUM OF PARTIAL CROSS SECTIONS 2637 1451 55 TAKEN FROM JENDL-3.2/1/. 2637 1451 56 2637 1451 57 MT = 102 CAPTURE CROSS SECTION 2637 1451 58 THE CAPTURE CROSS SECTION WAS ADOPTED FROM JENDL-3.2/1/. 2637 1451 59 2637 1451 60 IN THE RESONANCE REGION BELOW 350 KEV, THE CROSS SECTION WAS 2637 1451 61 CALCULATED FROM THE MULTILEVEL BREIT-WIGNER FORMULA WITH 2637 1451 62 RESONANCE PARAMETERS DETERMINED ON THE BASIS OF DATA BY GARG ET 2637 1451 63 AL./2/, KAEPPELER ET AL./3/, AND ALLEN AND MACKLIN/4/. ABOVE 2637 1451 64 350 KEV, THE CROSS SECTION WAS CALCULATED WITH THE OPTICAL AND 2637 1451 65 STATISTICAL MODEL CODE CASTHY/5/ BY NORMALIZING TO 3 MB AT 500 2637 1451 66 KEV/6/. CONTRIBUTIONS TO THE INELASTIC SCATTERING PROCESSES 2637 1451 67 WERE CONSIDERED BY ADOPTING THE EXCITED LEVELS IN THE FOLLOWING 2637 1451 68 TABLE. AND THE CONTRIBUTIONS TO THE (N,2N), (N,N'A), (N,N'P), 2637 1451 69 (N,P) AND (N,A) REACTION CROSS SECTIONS WERE ALSO CONSIDERED IN 2637 1451 70 THE CASTHY CALCULATION. THESE THRESHOLD REACTION CROSS SECTIONS 2637 1451 71 WERE CALCULATED WITH GNASH/7/. THE DIRECT CAPTURE CROSS SECTION 2637 1451 72 WAS CALCULATED WITH A SIMPLE FORMULA DERIVED BY BENZI AND 2637 1451 73 REFFO/8/ AND ADDED TO THE RESULT OF CASTHY CALCULATION. 2637 1451 74 2637 1451 75 OPTICAL POTENTIAL PARAMETERS 2637 1451 76 ----------------------------- 2637 1451 77 V = 46.0-0.25*EN (MEV), 2637 1451 78 WS = 14.0-0.2*EN (MEV), (IN THE GAUSSIAN FORM) 2637 1451 79 WI = 0.125*E-0.0004*E**2 (MEV), 2637 1451 80 VSO= 6.0 (MEV), 2637 1451 81 R = 1.286 (FM), A0 = 0.62 (FM) 2637 1451 82 RS = 1.387 (FM), AS = 0.7 (FM) 2637 1451 83 RSO= 1.07 (FM), ASO= 0.62 (FM) 2637 1451 84 2637 1451 85 LEVEL SCHEME 2637 1451 86 ------------ 2637 1451 87 NO. ENERGY(MEV) SPIN-PARITY 2637 1451 88 G.S. 0.0 0 + 2637 1451 89 1. 0.8108 2 + 2637 1451 90 2. 1.6747 2 + 2637 1451 91 3. 2.0765 4 + 2637 1451 92 4. 2.1339 3 + 2637 1451 93 5. 2.2581 0 + 2637 1451 94 6. 2.6004 4 + 2637 1451 95 7. 2.7819 1 + 2637 1451 96 8. 2.8764 2 + 2637 1451 97 9. 3.0840 2 + 2637 1451 98 10. 3.1330 4 + 2637 1451 99 11. 3.2330 2 + 2637 1451 100 12. 3.2440 0 + 2637 1451 101 LEVELS ABOVE 3.389 MEV WERE ASSUMED TO BE OVERLAPPING. 2637 1451 102 2637 1451 103 MF = 33 COVARIANCES OF NEUTRON CROSS SECTIONS 2637 1451 104 MT=102 CAPTURE CROSS SECTION 2637 1451 105 2637 1451 106 IN THE RESONANCE REGION, THE VARIANCE OF CROSS SECTION WAS 2637 1451 107 ESTIMATED FROM STANDARD DEVIATIONS OF THE RESONANCE PARAMETERS. 2637 1451 108 THE STANDARD DEVIATIONS WERE TAKEN FROM THE EXPERIMENTAL DATA 2637 1451 109 WHICH WERE CONSIDERED IN THE PARAMETER EVALUATION. FOR THE 2637 1451 110 RESONANCE PARAMETERS DETERMINED FROM THE CAPTURE AREA, THEIR 2637 1451 111 ERRORS WERE ESTIMATED FROM THE ERROR OF CAPTURE AREA. AN ERROR 2637 1451 112 OF 10 % WAS ASSUMED FOR OTHER CASES. 2637 1451 113 2637 1451 114 ABOVE 350 KEV, COVARIANCE MATRIX WAS OBTAINED BY USING 2637 1451 115 KALMAN/10/. ERRORS WERE CONSIDERED TO THE OPTICAL POTENTIAL AND 2637 1451 116 LEVEL DENSITY PARAMETERS, AND THE NORMALIZATION CROSS SECTION. 2637 1451 117 2637 1451 118 2637 1451 119 REFERENCES 2637 1451 120 1) NAKAGAWA T. ET AL.: J. NUCL. SCI. TECHNOL., 32, 1259 (1995). 2637 1451 121 2) GARG J.B. ET AL.: PHYS. REV., C18, 1141 (1978). 2637 1451 122 3) KAEPPELER F. ET AL.: NUCL. SCI. ENG., 84, 234 (1983). 2637 1451 123 4) ALLEN B.J. AND MACKLIN R.L.: J. PHYS. G., 6, 381 (1980). 2637 1451 124 5) IGARASI S. AND FUKAHORI T.: JAERI 1321 (1991). 2637 1451 125 6) TROFIMOV JU.N.: ATOMNAJA ENERGIJA, 58, 278 (1985). 2637 1451 126 7) YOUNG P.G. AND ARTHUR E.D.: LA-6974 (1977). 2637 1451 127 8) BENZI V. AND REFFO G.: CCDN/NW/10 (1969). 2637 1451 128 9) YAMAKOSHI H.: JAERI 1261, P.30 (1979). 2637 1451 129 10) KAWANO T. AND SHIBATA K.: JAERI-DATA/CODE 97-037 (1997) 2637 1451 130 [IN JAPANESE] 2637 1451 131 2637 1451 132 ******************************************************************2637 1451 133 The Q values and threshold energies were updated prior to pro- 2637 1451 134 cessing through the codes to comply with the values obtained 2637 1451 135 using the NNDC calculation program which is based on the 1995 2637 1451 136 Update to the Atomic mass Evaluation. 2637 1451 137 ******************************************************************2637 1451 138 ***************** Program LINEAR (VERSION 2002-1) ***************2637 1451 139 For All Data Greater than 1.0000E-10 barns in Absolute Value 2637 1451 140 Data Linearized to Within an Accuracy of .100000000 per-cent 2637 1451 141 ***************** Program RECENT (VERSION 2002-1) ***************2637 1451 142 for All Data Greater than 1.0000E-10 barns in Absolute Value 2637 1451 143 Data Linearized to within an Accuracy of .100000000 per-cent 2637 1451 144 ***************** Program SIGMA1 (VERSION 2002-1) ***************2637 1451 145 Data Doppler Broadened to 300.000000 Kelvin 2637 1451 146 for All Data Greater than 1.0000E-10 barns in Absolute Value 2637 1451 147 Data Linearized to Within an Accuracy pf .100000000 per-cent 2637 1451 148 ***************** Program FIXUP (Version 2002-1) ****************2637 1451 149 Corrected ZA/AWR in All Sections-----------------------------Yes 2637 1451 150 Corrected Thresholds-----------------------------------------Yes 2637 1451 151 Extended Cross Sections to 20 MeV----------------------------No 2637 1451 152 Allow Cross Section Deletion---------------------------------No 2637 1451 153 Allow Cross Section Reconstruction---------------------------No 2637 1451 154 Make All Cross Sections Non-Negative-------------------------Yes 2637 1451 155 Delete Energies Not in Ascending Order-----------------------Yes 2637 1451 156 Deleted Duplicate Points-------------------------------------Yes 2637 1451 157 Check for Ascending MAT/MF/MT Order--------------------------Yes 2637 1451 158 Check for Legal MF/MT Numbers--------------------------------Yes 2637 1451 159 Allow Creation of Missing Sections---------------------------No 2637 1451 160 Allow Insertion of Energy Points-----------------------------No 2637 1451 161 Create Uniform Energy Grid-----------------------------------No 2637 1451 162 Delete Section if Cross Section =0 at All Energies-----------Yes 2637 1451 163 ***************** Program GROUPIE (VERSION 2002-1) **************2637 1451 164 Unshielded Group Averages Using 640 Groups 2637 1451 165 Weighting Spectrum: Flat (Constant) Spectrum 2637 1451 166 1 451 170 12637 1451 167 2 151 4 12637 1451 168 3 102 217 12637 1451 169 33 102 200 12637 1451 170 2637 1 099999 2637 0 0 0 2.60580E+4 5.74356E+1 0 0 1 02637 2151 1 2.60580E+4 1.00000E+0 0 0 1 02637 2151 2 1.00000E-5 3.50000E+5 0 0 0 02637 2151 3 0.0 5.90000E-1 0 0 0 02637 2151 4 2637 2 099999 2637 0 0 0 2.60580E+4 5.74356E+1 0 0 0 02637 3102 1 6.58043E+6 6.58043E+6 0 0 1 6412637 3102 2 641 1 2637 3102 3 .000100000 20.4392055 .000105000 19.9625958 .000110000 19.50981102637 3102 4 .000115000 19.0925823 .000120000 18.6048355 .000127500 18.06369282637 3102 5 .000135000 17.5692785 .000142500 17.1149218 .000150000 16.62318192637 3102 6 .000160000 16.1109172 .000170000 15.6448196 .000180000 15.21697742637 3102 7 .000190000 14.8189064 .000200000 14.4554637 .000210000 14.11352532637 3102 8 .000220000 13.7959192 .000230000 13.5008306 .000240000 13.15445962637 3102 9 .000255000 12.7738323 .000270000 12.4801215 .000280000 12.15316722637 3102 10 .000300000 11.7544371 .000320000 11.3934012 .000340000 11.06275082637 3102 11 .000360000 10.7582019 .000380000 10.4801883 .000400000 10.19034412637 3102 12 .000425000 9.89374786 .000450000 9.62275629 .000475000 9.374305662637 3102 13 .000500000 9.14065477 .000525000 8.92711285 .000550000 8.725389682637 3102 14 .000575000 8.53778124 .000600000 8.34549314 .000630000 8.147677252637 3102 15 .000660000 7.96667299 .000690000 7.79303972 .000720000 7.608288232637 3102 16 .000760000 7.41019189 .000800000 7.22626354 .000840000 7.057702732637 3102 17 .000880000 6.89749502 .000920000 6.75062822 .000960000 6.609978432637 3102 18 .001000000 6.46419650 .001050000 6.31200628 .001100000 6.169327572637 3102 19 .001150000 6.03785946 .001200000 5.88277591 .001275000 5.712331392637 3102 20 .001350000 5.55649992 .001425000 5.41106895 .001500000 5.256562372637 3102 21 .001600000 5.09500822 .001700000 4.94752804 .001800000 4.811128322637 3102 22 .001900000 4.68651763 .002000000 4.57098309 .002100000 4.462571142637 3102 23 .002200000 4.36328773 .002300000 4.26834950 .002400000 4.160027262637 3102 24 .002550000 4.03966313 .002700000 3.94551646 .002800000 3.843217652637 3102 25 .003000000 3.71728764 .003200000 3.60269027 .003400000 3.497892912637 3102 26 .003600000 3.40230772 .003800000 3.31405188 .004000000 3.222064332637 3102 27 .004250000 3.12866520 .004500000 3.04342270 .004750000 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155 1.049430-3 1.048230-3 8.740180-4 6.452080-4 4.449160-4 3.301910-4263733102 156 2.914300-4 3.070620-4 5.609440-4 1.215170-3 1.734790-3 2.319580-3263733102 157 3.155040-3 3.065710-3 6.300100-4 6.532220-4 7.524960-4 7.617110-4263733102 158 7.936020-4 6.772240-4 7.573830-4 7.292780-4 5.352740-4 5.011720-4263733102 159 5.669470-4 7.002240-4 8.556530-4 9.622720-4 1.032640-3 1.042960-3263733102 160 1.020410-3 9.934750-4 9.421310-4 8.841950-4 8.629140-4 9.120660-4263733102 161 7.507600-4 7.684990-4 7.962920-4 8.833210-4 4.307810-4 7.942380-4263733102 162 7.291470-4 3.649350-4 3.326150-4 4.859060-4 7.565710-4 1.033560-3263733102 163 1.226240-3 1.328120-3 1.329670-3 1.178640-3 8.324920-4 5.325110-4263733102 164 1.729720-4-2.437260-4-1.493790-4 1.608010-3 1.626470-3 1.640970-3263733102 165 1.452110-3 1.526050-3 1.497130-3 1.653870-3 1.591680-3 1.626360-3263733102 166 1.745200-3 1.865800-3 1.942930-3 1.967390-3 1.888840-3 1.747240-3263733102 167 1.496650-3 1.273130-3 1.079570-3 1.216020-3 2.115460-3 1.755610-3263733102 168 1.771520-3 1.369290-3 1.601530-3 1.556760-3 1.699100-3 1.631830-3263733102 169 1.746660-3 1.934560-3 2.127160-3 2.235890-3 2.278390-3 2.201960-3263733102 170 1.900010-3 1.288550-3 7.795860-4 2.641850-4 0.000000+0 9.788250-4263733102 171 1.962490-3 1.117600-3 1.693930-3 1.608710-3 1.609390-3 1.572910-3263733102 172 1.792730-3 2.158490-3 2.513250-3 2.720840-3 2.784210-3 2.693820-3263733102 173 2.221200-3 1.039010-3 9.561710-5-8.607330-4-1.822950-3-9.269570-4263733102 174 7.089630-3 1.465430-3 1.637290-3 2.893500-3 2.716890-3 2.003500-3263733102 175 1.195020-3 5.785340-4 3.649000-4 4.572150-4 7.079900-4 1.750060-3263733102 176 4.057770-3 5.866950-3 7.838270-3 1.067520-2 1.038370-2 1.780840-3263733102 177 1.792260-3 2.161630-3 2.263660-3 2.346080-3 2.495750-3 2.601530-3263733102 178 2.643620-3 2.603900-3 2.520760-3 2.290590-3 1.852580-3 1.554630-3263733102 179 1.273120-3 9.360720-4 6.909550-4 1.962490-3 2.565350-3 2.718850-3263733102 180 2.738130-3 2.737270-3 2.706870-3 2.617010-3 2.543910-3 2.452930-3263733102 181 2.311910-3 2.188750-3 2.172640-3 2.191620-3 2.157160-3 1.411970-3263733102 182 4.664890-3 4.921800-3 4.558810-3 3.928810-3 3.214320-3 2.671990-3263733102 183 2.315360-3 2.166760-3 2.369560-3 3.283990-3 4.245070-3 5.342850-3263733102 184 6.652700-3 4.731700-3 5.685160-3 5.591200-3 5.274010-3 4.850020-3263733102 185 4.581470-3 4.495990-3 4.685950-3 5.286950-3 4.999010-3 5.436770-3263733102 186 6.162770-3 6.981340-3 4.208060-3 7.242010-3 8.144960-3 8.988140-3263733102 187 9.918580-3 1.092380-2 1.199030-2 1.333340-2 8.692290-3 6.805680-3263733102 188 5.439890-3 4.040780-3 8.571640-4 1.102500-2 1.306150-2 1.509890-2263733102 189 1.724460-2 1.918590-2 2.120880-2 1.205000-2 7.614630-3 3.937350-3263733102 190 1.748330-4-3.032670-3 1.716100-2 2.007790-2 2.296380-2 2.600090-2263733102 191 2.868950-2 1.526500-2 8.626110-3 2.794700-3-2.965560-3-6.111480-3263733102 192 2.496400-2 2.837000-2 3.205500-2 3.559640-2 1.889820-2 1.035410-2263733102 193 2.874230-3-4.179100-3-6.893270-3 3.422500-2 3.780380-2 4.258630-2263733102 194 2.280370-2 1.270380-2 4.104470-3-3.570120-3-5.771380-3 4.452100-2263733102 195 4.894050-2 2.666180-2 1.559110-2 6.239220-3-1.641400-3-3.268030-3263733102 196 5.904900-2 3.828800-2 2.968610-2 2.309590-2 2.096800-2 2.670880-2263733102 197 6.150400-2 7.269370-2 8.635430-2 1.136110-1 1.503490-1 1.004890-1263733102 198 1.229590-1 1.670490-1 2.212880-1 1.648360-1 2.211800-1 2.929660-1263733102 199 3.226240-1 4.279660-1 6.352090-1 263733102 200 263733 099999 2637 0 0 0 0 0 0 0 2.70590E+4 5.84269E+1 0 0 34 102725 1451 1 0.0 0.0 0 0 0 62725 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022725 1451 3 3.00000E+2 0.0 1 0 233 82725 1451 4 27-Co- 59 IRK/FEI EVAL-APR90 VONACH ET AL., K.I.Zolotarev 2725 1451 5 DIST-Feb2004 2725 1451 6 ----IRDF-2002 MATERIAL 2725 2725 1451 7 -----INCIDENT NEUTRON DATA 2725 1451 8 ------ENDF-6 FORMAT 2725 1451 9 ******************************************************************2725 1451 10 27-CO- 59 ANL,ORNL EVAL-JUL89 A.SMITH+,G.DESAUSSURE+ 2725 1451 11 ANL/NDM-107 DIST-JUN93 REV1-JUN92 19930930 2725 1451 12 ----ENDF/B-VI MATERIAL 2725 REVISION 1 2725 1451 13 ******************************************************************2725 1451 14 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2725 1451 15 DOSIMETRY 2725 1451 16 ******************************************************************2725 1451 17 ***************************************************************** 2725 1451 18 ********** Start of (N,G) bibliographical component *********** 2725 1451 19 MF=2,MT=151, MF=3,MT=1,2,102 taken from ENDF/B-VI Release 2 2725 1451 20 MF=33,MT=102 taken from IRDF-90 2725 1451 21 ***************************************************************** 2725 1451 22 REVISION 1 -- NEW RESONANCE REGION EVALUATION BY G.DESAUSSURE, 2725 1451 23 N.M.LARSON, J.A.HARVEY, N.W.HILL, ANN. NUCL. ENERGY 19, 393 2725 1451 24 (1992) 2725 1451 25 ORIGINAL EVALUATION 2725 1451 26 A.SMITH,D.SMITH,P.GUENTHER,J.MEADOWS, R.LAWSON(ANL), 2725 1451 27 R.HOWERTON(LLNL),M.SUGIMOTO(JAERI) 2725 1451 28 ******************************************************************2725 1451 29 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 2725 1451 30 * * * * * * *2725 1451 31 R.LAWSON(ANL),R.HOWERTON(LLNL),M.SUGIMOTO(JAERI) 2725 1451 32 ***************************************************************** 2725 1451 33 ********** End of (N,G) bibliographical component *********** 2725 1451 34 ***************************************************************** 2725 1451 35 ***************************************************************** 2725 1451 36 ********** Start of (N,2N) bibliographical component ********** 2725 1451 37 ***************************************************************** 2725 1451 38 ++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++2725 1451 39 MF/MT 2151, MF/MT 3 16 AND MF/MT 33 16 FROM FOLLOWING 2725 1451 40 EVALUATION 2725 1451 41 27-CO- 59 IRK-VIENNA EVAL-APR90 2725 1451 42 DIST-JUN90 2725 1451 43 IRK-EVAL.NLIB 25 2725 2725 1451 44 ++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++2725 1451 45 ***************************************************************** 2725 1451 46 ********** End of (N,2N) bibliographical component *********** 2725 1451 47 ***************************************************************** 2725 1451 48 ***************************************************************** 2725 1451 49 ********** Start of (N,A) bibliographical component ********** 2725 1451 50 ***************************************************************** 2725 1451 51 ***************************************************************** 2725 1451 52 ------Russian Reactor Dosimetry File RRDF-2002 2725 1451 53 ***************************************************************** 2725 1451 54 Author of evaluation: K.I.Zolotarev 2725 1451 55 ***************************************************************** 2725 1451 56 MF=3 2725 1451 57 MT=107 -(n,a) cross section 2725 1451 58 ------------------------------------- 2725 1451 59 Excitation function for the Co-59(n,a)Mn-56 reaction in the 2725 1451 60 energy region from 2.5 MeV to 21 MeV was evaluated by means of 2725 1451 61 statistical analysis of experimental cross section data [1-28], 2725 1451 62 and data from GNASH [29] calculation. 2725 1451 63 All experimental data were renormalized to the new standards 2725 1451 64 for monitor reactions cross sections and decay data. Uncertainty 2725 1451 65 in the monitor reaction cross section was added to the total 2725 1451 66 uncertainty for Santry and Butler data [5]. Cross section data 2725 1451 67 measured by Huang Jianzhou et al. [18] were renormalized to the 2725 1451 68 preliminary evaluated integral of excitation function in the 2725 1451 69 energy range 13.0 - 15.0 MeV. Correction factor for this data 2725 1451 70 was equal Fc= 1.0507 . 2725 1451 71 Experimental data from ref.[30-41] were rejected due to their 2725 1451 72 discrepancy with the main bulk of experimental data [1-28] and 2725 1451 73 data from theoretical model calculation. 2725 1451 74 The final procedure of evaluation Co-59(n,a)Mn-56 excitation 2725 1451 75 function from threshold to 21 MeV has been carried out within the 2725 1451 76 framework of generalized least squares method. Rational function 2725 1451 77 was used as model function [42]. Calculations was performed by 2725 1451 78 means of Pade-2 code [43]. 2725 1451 79 The evaluated Co-59(n,a)Mn-56 excitation function averaged 2725 1451 80 on U-235 neutron fission spectrum [44] and Cf-252 spontaneous 2725 1451 81 fission neutron spectrum [45] gives the next values : 2725 1451 82 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2725 1451 83 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2725 1451 84 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2725 1451 85 ³ ³ 0.1500 +- 0.0080 [46] 2725 1451 86 U-235 neutron fission ³ 0.15823 ³ 0.1700 +- 0.0120 [47] 2725 1451 87 ³ ³ 0.1568 +- 0.0035 [48] 2725 1451 88 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2725 1451 89 ³ ³ 0.2208 +- 0.0014 [49] 2725 1451 90 Cf-252 spont. fission ³ 0.22095 ³ 0.2221 +- 0.0023 [50] 2725 1451 91 ³ ³ 0.2218 +- 0.0041 [51] 2725 1451 92 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2725 1451 93 2725 1451 94 MT=33 2725 1451 95 MT=107 -(n,a) cross section cov. matrix 2725 1451 96 ------------------------------------------- 2725 1451 97 Uncertainties in the evaluated excitation function for the 2725 1451 98 reaction Co-59(n,a)Mn-56 are given in the form of relative cova- 2725 1451 99 riance matrix for the 32-neutron energy groups (LB=5). Covariance 2725 1451 100 matrix of uncertainties was calculated simultaneously with recom- 2725 1451 101 mended cross section data by means of PADE-2 code. 2725 1451 102 Eigenvalues of the 6-th digits relative covariance matrix 2725 1451 103 given in the 33-file are the following: 2725 1451 104 2725 1451 105 1.11275E-08 1.13832E-08 1.19424E-08 1.27401E-08 2725 1451 106 1.37409E-08 1.52188E-08 1.67720E-08 1.86812E-08 2725 1451 107 2.09005E-08 2.38172E-08 2.69210E-08 3.07595E-08 2725 1451 108 3.46853E-08 4.07806E-08 4.51537E-08 5.55085E-08 2725 1451 109 6.19014E-08 7.83507E-08 9.39864E-08 1.01329E-07 2725 1451 110 2.19793E-07 8.71735E-07 1.71048E-05 1.93185E-04 2725 1451 111 1.05065E-03 1.19481E-03 2.01786E-03 2.53733E-03 2725 1451 112 3.28491E-03 1.06037E-02 2.32281E-02 9.30559E-01 2725 1451 113 2725 1451 114 References : 2725 1451 115 1. H.G.Blosser et.al. Phys. Rev., v.110, p.531, 1958 2725 1451 116 2. I.L.Preiss,R.W.Fink Nucl. Phys., v.15, p.326, February 1960 2725 1451 117 3. C.S.Khurana, H.S.Hans Proc.of 4th Nuclear Physics and Solid 2725 1451 118 State Physics Symp., 24-26 Feb. 1960, Waltair, India, p.297 2725 1451 119 4. E.Weigold Australian J. Phys., v.13, p.186, 1960 2725 1451 120 5. D.C.Santry, J.P.Butler Can. J. Phys., v.42, p.1030, 1964 2725 1451 121 6. H.Liskien, A.Paulsen J. Nucl. of Energy, v.19, p.73, 1965 2725 1451 122 7. K.S.Khurana,I.M.Govil Nucl. Phys., v.69, p.153, July 1965 2725 1451 123 8. H.Liskien, A.Paulsen Nucleonics, v.8, p.315, June 1966 2725 1451 124 9. V.N.Levkovskiy et.al. Yadernaja Fizika (Sov.), v.8, n.1, p.7, 2725 1451 125 July 1968 2725 1451 126 10. J.C.Robertson, B.Audric, P.Kolkowski J. Nucl. Energ., v.27, 2725 1451 127 p.531, August 1973 2725 1451 128 11. S.K.Ghorai, J.E.Gaiser, W.L.Alford Annals of Nucl. Energy, 2725 1451 129 v.7, p.41, 1980 2725 1451 130 12. E.Zupranska et al. Progress Report INR-1821/I/PL/A, 1980 2725 1451 131 E.Zupranska et al. Acta. Phys. Pol., v.B11, p.853, Nov. 1980 2725 1451 132 13. B.M.Bahal, R.Pepelnik Report GKSS-84-E-, 1984 2725 1451 133 14. J.W.Meadows, D.L.Smith et al. Annals of Nucl. Energy, v.14, 2725 1451 134 p.489, 1987 2725 1451 135 15. J.W.Meadows, D.L.Smith, R.D.Lawson Annals of Nucl. Energy, 2725 1451 136 v.14, p.603, 1987 2725 1451 137 16. Y.Ikeda, C.Konno, K.Oishi et al. Report JAERI-1312, 1988 2725 1451 138 17. T.B.Ryves, P.Kolkowski, S.M.Judge Annals of Nuclear Energy, 2725 1451 139 v.15, p.561, Dec.1988 2725 1451 140 18. Huang Jianzhou et al. Report INDC(CPR)-16, Vienna,August 1989 2725 1451 141 19. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 2725 1451 142 20. Li Tingyan et al. High Energy Physics and Nuclear Physics, 2725 1451 143 v.14(6), p.542, June 1990 2725 1451 144 21. A.Ercan et al. Proc. of an Intern. Conf. on Nuclear Data for 2725 1451 145 Science and Technology, 13-17 May 1991, Julich, FRG, Springer-2725 1451 146 Verlag, 1992, p.376-377 2725 1451 147 22. Y.Ikeda et al. Progress Report INDC(JPN)-162/U,p.24, Aug.1992 2725 1451 148 23. A.Grallert et al. Report INDC(NDS)-286, p.193, IAEA, 1993 2725 1451 149 24. W.Mannhart et al. Proc.of International Conference on Nuclear 2725 1451 150 Data for Science and Technology, Gatlinburg, Tennessee, May 2725 1451 151 9-13, 1994, v.1, pp.285-287 2725 1451 152 25. A.A.Filatenkov et al. VANT, Ser.:Yadernye Konstanty, v.2, p.8,2725 1451 153 Moscow, 1996 2725 1451 154 26. K.T.Osman, F.I.Habbani Report INDC(SUD)-001, Distr.L, IAEA, 2725 1451 155 Vienna, October 1996 2725 1451 156 27. A.D.Majdeddin Measurement and analysis of excitation func- 2725 1451 157 tions for fast neutron induced reactions from threshold to 2725 1451 158 20 MeV. PHD dissertation, Inst. of Experimental Physics , 2725 1451 159 Kossuth University, Hungary, June 1998 2725 1451 160 28. A.A.Filatenkov et al. Report RI-252, St. Petersburg, May 1999 2725 1451 161 29. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 2725 1451 162 Model Code for Calculation of Cross Section and Emission 2725 1451 163 Spectra. Report LA-6947, Los Alamos, 1977 ; 2725 1451 164 E.L.Trykiv, G.Ya.Tertychnyi Private communication, IPPE, 2725 1451 165 Obninsk, May 1999 2725 1451 166 30. E.B.Paul,R.L.Clarke Canad. J. of Phys., v.31, p.267, 1953 2725 1451 167 31. M.Bormann et al. Journal de Physique, v.22, p.602, Oct. 1961 2725 1451 168 32. F.Gabbard, B.D.Kern Phys. Rev., v.128, n.3, p.1276, 1962 2725 1451 169 33. J.M.F.Jeronymo et al. Nucl. Phys., v.47(1), p.157, July 1963 2725 1451 170 34. J.E.Strain, W.J.Ross Report ORNL-3672, January 1965 2725 1451 171 35. M.Bormann et al. Nucl. Phys., v.63, p.438, March 1965 2725 1451 172 36. U.Garuska, J.Dresler, H.Malecki Prog. Report INR-1773/I/PL/A,2725 1451 173 p.16, September 1978 2725 1451 174 37. H.M.Agrawal, K.R.Zasadny, G.F.Knoll Trans. Amer. Nucl. Soc., 2725 1451 175 v.47, p.431, November 1984 save 2725 1451 176 38. M.Berrada Progress Report to the IAEA NDS on Research 2725 1451 177 Contract -3311.R1/RB, 15 April 1984 2725 1451 178 39. I.Garlea et al. Revue Roumaine de Physique, v.30, p.673, 1985 2725 1451 179 40. N.I.Molla et al. Proc.of International Conference on Nuclear 2725 1451 180 Data for Science and Technology, Gatlinburg, Tennessee, May 2725 1451 181 9-13, 1994, v.2, pp.938-940 2725 1451 182 41. R.Doczi et al. Nucl. Sci. Eng., v.129, p.164, June 1998 2725 1451 183 42. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2725 1451 184 st's Meeting on Evaluation and Processing of Covariance Data, 2725 1451 185 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 2725 1451 186 43. S.A.Badikov et.al. Preprint FEI-1686, Obninsk, 1985 2725 1451 187 44. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2725 1451 188 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2725 1451 189 45. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2725 1451 190 46. K.Kobayashi, I.Kimura Report NEANDC(J)-61U, p.81, Sep. 1979 2725 1451 191 47. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2725 1451 192 Washington D.C., 25-28 April 1989, v.2, p.923 2725 1451 193 48. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 2725 1451 194 49. K.Kobayashi,I.Kimura,W.Mannhart Nucl. Sci. Technology.,v.19, 2725 1451 195 n.5, p.341, May 1982 2725 1451 196 50. W.Mannhart Handbook on Nuclear Activation Data , IAEA 2725 1451 197 Technical Report series No.273, p.413, 1987 2725 1451 198 51. W.Mannhart Validation of Differential Cross Sections with 2725 1451 199 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2725 1451 200 September 2002 2725 1451 201 ***************************************************************** 2725 1451 202 ********** End of (N,A) bibliographical component ********** 2725 1451 203 ***************************************************************** 2725 1451 204 The Q values and threshold energies were updated prior to pro- 2725 1451 205 cessing through the codes to comply with the values obtained 2725 1451 206 using the NNDC calculation program which is based on the 1995 2725 1451 207 Update to the Atomic mass Evaluation. 2725 1451 208 ***************************************************************** 2725 1451 209 ***************** Program LINEAR (VERSION 2002-1) ***************2725 1451 210 For All Data Greater than 1.0000E-10 barns in Absolute Value 2725 1451 211 Data Linearized to Within an Accuracy of .100000000 per-cent 2725 1451 212 ***************** Program RECENT (VERSION 2002-1) ***************2725 1451 213 for All Data Greater than 1.0000E-10 barns in Absolute Value 2725 1451 214 Data Linearized to within an Accuracy of .100000000 per-cent 2725 1451 215 ***************** Program SIGMA1 (VERSION 2002-1) ***************2725 1451 216 Data Doppler Broadened to 300.000000 Kelvin 2725 1451 217 for All Data Greater than 1.0000E-10 barns in Absolute Value 2725 1451 218 Data Linearized to Within an Accuracy pf .100000000 per-cent 2725 1451 219 ***************** Program FIXUP (Version 2002-1) ****************2725 1451 220 Corrected ZA/AWR in All Sections-----------------------------Yes 2725 1451 221 Corrected Thresholds-----------------------------------------Yes 2725 1451 222 Extended Cross Sections to 20 MeV----------------------------No 2725 1451 223 Allow Cross Section Deletion---------------------------------No 2725 1451 224 Allow Cross Section Reconstruction---------------------------No 2725 1451 225 Make All Cross Sections Non-Negative-------------------------Yes 2725 1451 226 Delete Energies Not in Ascending Order-----------------------Yes 2725 1451 227 Deleted Duplicate Points-------------------------------------Yes 2725 1451 228 Check for Ascending MAT/MF/MT Order--------------------------Yes 2725 1451 229 Check for Legal MF/MT Numbers--------------------------------Yes 2725 1451 230 Allow Creation of Missing Sections---------------------------No 2725 1451 231 Allow Insertion of Energy Points-----------------------------No 2725 1451 232 Create Uniform Energy Grid-----------------------------------No 2725 1451 233 Delete Section if Cross Section =0 at All Energies-----------Yes 2725 1451 234 ***************** Program GROUPIE (VERSION 2002-1) **************2725 1451 235 Unshielded Group Averages Using 640 Groups 2725 1451 236 Weighting Spectrum: Flat (Constant) Spectrum 2725 1451 237 1 451 245 12725 1451 238 2 151 4 12725 1451 239 3 16 35 12725 1451 240 3 102 217 12725 1451 241 3 107 75 12725 1451 242 33 16 42 12725 1451 243 33 102 11 12725 1451 244 33 107 103 12725 1451 245 2725 1 099999 2725 0 0 0 2.70590E+4 5.84269E+1 0 0 1 02725 2151 1 2.70590E+4 1.00000E+0 0 0 1 02725 2151 2 1.00000E-5 1.00000E+5 0 0 0 02725 2151 3 3.50000E+0 6.67200E-1 0 0 0 02725 2151 4 2725 2 099999 2725 0 0 0 2.70590E+4 5.84269E+1 0 0 0 02725 3 16 1 -1.04530E+7-1.04530E+7 0 0 1 952725 3 16 2 95 1 2725 3 16 3 10600000.0 .001438929 10700000.0 .007329788 10800000.0 .0135365962725 3 16 4 10900000.0 .026794286 11000000.0 .047102857 11100000.0 .0674114292725 3 16 5 11200000.0 .088045429 11300000.0 .110632000 11400000.0 .1335440002725 3 16 6 11500000.0 .156456000 11600000.0 .179368000 11700000.0 .2030660002725 3 16 7 11800000.0 .231480000 11900000.0 .260680000 12000000.0 .2898800002725 3 16 8 12100000.0 .319080000 12200000.0 .348198750 12300000.0 .3768300002725 3 16 9 12400000.0 .405380000 12500000.0 .433930000 12600000.0 .4624800002725 3 16 10 12700000.0 .489566125 12800000.0 .507869000 12900000.0 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1.424180-3 1.114220-3 7.625800-4272533107 35 5.617700-4 5.141090-4 5.430070-4 5.820970-4 5.979410-4 5.813950-4272533107 36 5.358150-4 4.698660-4 3.941360-4 3.194570-4 2.555220-4 2.092830-4272533107 37 1.835460-4 1.765940-4 1.832940-4 1.971400-4 2.122100-4 2.242630-4272533107 38 2.309360-4 2.313760-4 2.257250-4 2.146560-4 1.990550-4 1.798340-4272533107 39 1.578280-4 1.067160-3 8.670970-4 6.654690-4 5.383210-4 4.857530-4272533107 40 4.779590-4 4.861430-4 4.912880-4 4.831270-4 4.575070-4 4.146560-4272533107 41 3.583800-4 2.954790-4 2.346000-4 1.842730-4 1.507000-4 1.363490-4272533107 42 1.400090-4 1.580380-4 1.859100-4 2.194060-4 2.551900-4 2.909270-4272533107 43 3.251460-4 3.570250-4 3.861870-4 4.125400-4 8.671810-4 7.714650-4272533107 44 6.471120-4 5.449710-4 4.815700-4 4.514700-4 4.413730-4 4.375690-4272533107 45 4.287860-4 4.072840-4 3.696310-4 3.173050-4 2.565290-4 1.967090-4272533107 46 1.476230-4 1.165570-4 1.067360-4 1.175070-4 1.456570-4 1.869280-4272533107 47 2.371300-4 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60 8.478330-4 8.276790-4 7.647320-4 6.738120-4 5.686930-4 4.617520-4272533107 61 3.639960-4 2.845050-4 2.291860-4 1.995590-4 1.925450-4 2.017460-4272533107 62 2.196390-4 2.395840-4 2.569040-4 2.690010-4 2.749150-4 2.747480-4272533107 63 2.691630-4 2.590450-4 2.453050-4 2.287740-4 8.395720-4 8.040470-4272533107 64 7.322570-4 6.362260-4 5.281700-4 4.203680-4 3.245530-4 2.503860-4272533107 65 2.035030-4 1.842990-4 1.884050-4 2.086620-4 2.374990-4 2.686370-4272533107 66 2.977680-4 3.224500-4 3.416560-4 3.552660-4 3.636700-4 3.674940-4272533107 67 3.674300-4 8.006990-4 7.588770-4 6.856440-4 5.899000-4 4.825730-4272533107 68 3.762830-4 2.838360-4 2.156520-4 1.772910-4 1.685220-4 1.844370-4272533107 69 2.177930-4 2.612660-4 3.088450-4 3.562780-4 4.009380-4 4.414360-4272533107 70 4.772250-4 5.082750-4 5.348470-4 7.504010-4 7.072690-4 6.334420-4272533107 71 5.366440-4 4.286200-4 3.239150-4 2.369800-4 1.785680-4 1.532970-4272533107 72 1.595740-4 1.915490-4 2.416910-4 3.028130-4 3.691110-4 4.364210-4272533107 73 5.020320-4 5.643450-4 6.225410-4 6.763130-4 6.954900-4 6.488650-4272533107 74 5.711140-4 4.716320-4 3.648490-4 2.673910-4 1.938160-4 1.529760-4272533107 75 1.467730-4 1.714680-4 2.202600-4 2.856770-4 3.610820-4 4.412890-4272533107 76 5.225880-4 6.025210-4 6.795740-4 7.529000-4 6.304990-4 5.779750-4272533107 77 4.974380-4 4.012420-4 3.056610-4 2.266320-4 1.755580-4 1.571880-4272533107 78 1.701860-4 2.093370-4 2.679510-4 3.395610-4 4.187810-4 5.015220-4272533107 79 5.848930-4 6.669610-4 7.465100-4 5.538450-4 5.011350-4 4.285190-4272533107 80 3.492380-4 2.776720-4 2.253930-4 1.986950-4 1.984520-4 2.216480-4272533107 81 2.633770-4 3.184200-4 3.821410-4 4.508370-4 5.217500-4 5.929330-4272533107 82 6.630700-4 4.818940-4 4.435680-4 3.942170-4 3.438620-4 3.016870-4272533107 83 2.738300-4 2.627190-4 2.677410-4 2.864790-4 3.158430-4 3.528180-4272533107 84 3.948190-4 4.398050-4 4.862480-4 5.330440-4 4.454760-4 4.358260-4272533107 85 4.183730-4 3.977630-4 3.782510-4 3.628080-4 3.529130-4 3.488350-4272533107 86 3.500810-4 3.558030-4 3.650640-4 3.769920-4 3.908400-4 4.060010-4272533107 87 4.694740-4 4.919110-4 5.021020-4 5.010390-4 4.910640-4 4.749580-4272533107 88 4.552730-4 4.340100-4 4.125720-4 3.918450-4 3.723330-4 3.542780-4272533107 89 3.377530-4 5.551840-4 6.013500-4 6.274720-4 6.341600-4 6.243770-4272533107 90 6.020350-4 5.709710-4 5.344320-4 4.949130-4 4.542130-4 4.135620-4272533107 91 3.737590-4 6.840580-4 7.432410-4 7.770160-4 7.873120-4 7.781630-4272533107 92 7.542020-4 7.197360-4 6.783370-4 6.327650-4 5.850490-4 5.366280-4272533107 93 8.380390-4 9.069060-4 9.497230-4 9.693530-4 9.700350-4 9.561740-4272533107 94 9.317100-4 8.998770-4 8.631920-4 8.235440-4 1.015840-3 1.100660-3272533107 95 1.162040-3 1.202680-3 1.226130-3 1.235970-3 1.235370-3 1.226980-3272533107 96 1.212910-3 1.233620-3 1.346000-3 1.438180-3 1.512010-3 1.570060-3272533107 97 1.614950-3 1.649110-3 1.674650-3 1.515040-3 1.666300-3 1.799460-3272533107 98 1.915510-3 2.016160-3 2.103310-3 2.178850-3 1.880800-3 2.078470-3272533107 99 2.258530-3 2.421410-3 2.568290-3 2.700660-3 2.343020-3 2.589960-3272533107 100 2.818450-3 3.028720-3 3.221760-3 2.904570-3 3.199650-3 3.474650-3272533107 101 3.729920-3 3.560920-3 3.900280-3 4.217700-3 4.302840-3 4.681140-3272533107 102 5.118740-3 272533107 103 272533 099999 2725 0 0 0 0 0 0 0 2.80580E+4 5.74376E+1 0 0 34 102825 1451 1 0.0 0.0 0 0 0 62825 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022825 1451 3 3.00000E+2 0.0 1 0 378 62825 1451 4 28-Ni- 58 IRK-IJS/FEIEVAL-AUG99 EUR.JOINT COLLAB. K.I.Zolotarev 2825 1451 5 DIST-Feb2004 2825 1451 6 ----IRDF-2002 MATERIAL 2825 2825 1451 7 -----INCIDENT NEUTRON DATA 2825 1451 8 ------ENDF-6 FORMAT 2825 1451 9 2825 1451 10 ***************************** JEFF-3.0 ***********************2825 1451 11 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2825 1451 12 DOSIMETRY 2825 1451 13 ******************************************************************2825 1451 14 28-NI- 58 IRK-IJS EVAL-AUG99 EUROPEAN JOINT COLLABORATION 2825 1451 15 2825 1451 16 DATA TAKEN FROM :- EFF-3.1 (DIST-AUG99 REV1-SEP00) 2825 1451 17 MF=3 2825 1451 18 MT= 16 - (n,2n) 2825 1451 19 2825 1451 20 ******************************************************************2825 1451 21 28-Ni- 58 FEI EVAL-Dec02 K.I.Zolotarev 2825 1451 22 DIST-Jan03 2825 1451 23 ----BROND-3 MATERIAL 2825 2825 1451 24 -----INCIDENT NEUTRON DATA 2825 1451 25 ------ENDF-6 FORMAT 2825 1451 26 ***************************************************************** 2825 1451 27 ------Russian Reactor Dosimetry File RRDF-2002 2825 1451 28 ***************************************************************** 2825 1451 29 Author of evaluation: K.I.Zolotarev 2825 1451 30 ***************************************************************** 2825 1451 31 MF=3 2825 1451 32 MT=103 - Ni58(n,p)Co58m+g reaction 2825 1451 33 2825 1451 34 ***************************************************************** 2825 1451 35 ******* Start of JEFF-3.0 (N,2N) bibliographical component ****** 2825 1451 36 ***************************************************************** 2825 1451 37 ***************************** JEFF-3.0 ***********************2825 1451 38 28-NI- 58 IRK-IJS EVAL-AUG99 EUROPEAN JOINT COLLABORATION 2825 1451 39 2825 1451 40 DATA TAKEN FROM :- EFF-3.1 (DIST-AUG99 REV1-SEP00) 2825 1451 41 2825 1451 42 ******************************************************************2825 1451 43 Authors and Responsibilities: 2825 1451 44 S.Tagesen, H.Vonach and A.Wallner, I.R.K.: 2825 1451 45 - Complete evaluation of the cross sections including covariance 2825 1451 46 matrices by generalized least squares cross section update 2825 1451 47 code GLUCS (9, 10). The total cross section is evaluated in 2825 1451 48 broad energy bins. 2825 1451 49 A.Trkov, I.J.S.: 2825 1451 50 - Final assembly of the file. 2825 1451 51 - Consistency corrections (interp.law in MF6 starter file). 2825 1451 52 - Implementation of the resonance fluctuations on the smooth 2825 1451 53 newly evaluated cross sections. 2825 1451 54 - Preliminary data verification and benchmarking. 2825 1451 55 2825 1451 56 Evaluation Details: 2825 1451 57 The Oak Ridge Ni-58 ENDF/B-VI Revision 1 evaluation (MAT 2825) 2825 1451 58 by Larson et.al. was the chosen starter file. All neutron 2825 1451 59 cross sections above 810 keV were re-evaluated and include 2825 1451 60 covariance data. The following data sets were selected as 2825 1451 61 "priors": 2825 1451 62 MT16 IRDF-90 (evaluation Pavlik)(11) 2825 1451 63 2825 1451 64 Resonance Fluctuations in the Cross Sections: 2825 1451 65 The cross sections below 810 keV are not affected by the re- 2825 1451 66 evaluation process. Above this energy, broad bin average total 2825 1451 67 cross section from the starter file was calculated. The bins 2825 1451 68 correspond to those used in the re-evaluation process. A smooth 2825 1451 69 cross section curve was generated, conserving the bin average 2825 1451 70 values. Fluctuations modulating function was defined as the 2825 1451 71 ratio of the original and the smoothed cross section. This 2825 1451 72 modulating function is applied on the re-evaluated smooth total 2825 1451 73 cross section and the inelastic cross sections. 2825 1451 74 2825 1451 75 The remaining comments are taken over from the starter file, 2825 1451 76 except for the sections referring to the data, which have been 2825 1451 77 superseeded. Modified sections are identified by the use of 2825 1451 78 lower case characters. 2825 1451 79 2825 1451 80 ******************************************************************2825 1451 81 CAPTURE WIDTHS CORRECTED FOR 58.7 AND 439.52 KEV RESONANCES. 2825 1451 82 THE ELASTIC TRANSFORMATION MATRIX WAS REMOVED. 2825 1451 83 FIXED TYPO IN MINIMUM ENERGY IN MF=6, MT=51 2825 1451 84 ******************************************************************2825 1451 85 2825 1451 86 THIS WORK EMPLOYED NUCLEAR MODEL CODES INCLUDING THE 2825 1451 87 DISTORTED WAVE BORN APPROXIMATION (DWBA) PROGRAM DWUCK (1) 2825 1451 88 AND THE HAUSER-FESHBACH CODE TNG (2,3,4). THE TNG CODE PROVIDES 2825 1451 89 ENERGY AND ANGULAR DISTRIBUTIONS OF PARTICLES EMITTED IN THE 2825 1451 90 COMPOUND AND PRE-COMPOUND REACTIONS, ENSURES CONSISTENCY AMONG ALL2825 1451 91 REACTIONS, AND MAINTAINS ENERGY BALANCE. DETAILS PERTINENT TO THE 2825 1451 92 CONTENTS OF THIS EVALUATION AND EXTENSIVE COMPARISONS OF 2825 1451 93 CALCULATIONS WITH EXPERIMENTAL DATA CAN BE FOUND IN REFERENCE (5).2825 1451 94 2825 1451 95 ----- DESCRIPTION OF FILES 2825 1451 96 (MF-MT) 2825 1451 97 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2825 1451 98 3-16 (n,2n) cross sections were re-evaluated. 2825 1451 99 -------------------------------------------------------------- 2825 1451 100 UNCERTAINTY FILES 2825 1451 101 ALL NON-DERIVED FILES CONTAIN AN LB=8 COMPONENT, AS 2825 1451 102 REQUIRED BY ENDF/B-VI FORMATS 2825 1451 103 For all evaluated reactions full covariance matrices are 2825 1451 104 given as calculated by the bayesian evaluation update code2825 1451 105 GLUCS. This includes full inter-reaction covariances. 2825 1451 106 2825 1451 107 33-16 (N,2N) covariances from GLUCS. 2825 1451 108 2825 1451 109 REFERENCES: 2825 1451 110 (11) M. Wagner et al., Physics Data 13-5, Fachinformationszentrum 2825 1451 111 Karlsruhe, 1990 2825 1451 112 ***************************************************************** 2825 1451 113 ******* End of JEFF-3.0 (N,2N) bibliographical component ******* 2825 1451 114 ***************************************************************** 2825 1451 115 ***************************************************************** 2825 1451 116 ******* Start of RRDF (N,P) bibliographical component ******* 2825 1451 117 ***************************************************************** 2825 1451 118 28-Ni- 58 FEI EVAL-Dec02 K.I.Zolotarev 2825 1451 119 DIST-Jan03 2825 1451 120 ----BROND-3 MATERIAL 2825 2825 1451 121 -----INCIDENT NEUTRON DATA 2825 1451 122 ------ENDF-6 FORMAT 2825 1451 123 ***************************************************************** 2825 1451 124 ------Russian Reactor Dosimetry File RRDF-2002 2825 1451 125 ***************************************************************** 2825 1451 126 Author of evaluation: K.I.Zolotarev 2825 1451 127 ***************************************************************** 2825 1451 128 MF=3 2825 1451 129 MT=103 - Ni58(n,p)Co58m+g reaction 2825 1451 130 ------------------------------------- 2825 1451 131 Microscopic experimental data [1-64] were analyzed in the 2825 1451 132 process of preparation of input data base for the evaluation of 2825 1451 133 cross sections and their uncertainty for the Ni-58(n,p)Co-58m+g 2825 1451 134 reaction. During this procedure all experimental data if it was 2825 1451 135 possible were corrected to the new recommended cross section data 2825 1451 136 for monitor reactions used in the measurements and to the new re- 2825 1451 137 commended decay data. 2825 1451 138 Excitation function for the Ni-58(n,p)Co-58m+g reaction in 2825 1451 139 the energy region from threshold to 20 MeV was evaluated by means 2825 1451 140 of statistical analysis of experimental cross section data [1-41].2825 1451 141 Special correction was done with experimental data [9], [12], 2825 1451 142 [13], [15], [16], [20], [23], [26], [34], [35]. Experimental data 2825 1451 143 of Decowski et al. [9] corresponding to the neutron energy region 2825 1451 144 12.42 - 17.08 MeV were renormalized to the preliminary evaluated 2825 1451 145 cross section value at En=14.77 MeV. Original cross sections from 2825 1451 146 ref. [9] were multiplied to the correction factor Fc=0.53807 . 2825 1451 147 Data of Paulsen and Widera [12] measured in the energy ranges 2825 1451 148 3.35 - 6.27 MeV and 12.8 - 16.4 MeV were corrected to the preli- 2825 1451 149 minary evaluated cross section value at 6.3 MeV and integral of 2825 1451 150 cross section in the energy interval 14-15 MeV. The correction 2825 1451 151 factors for the experimental data [12] were Fc=1.059 and Fc=1.058,2825 1451 152 respectively. 2825 1451 153 Data of Smith and Meadows [13] measured with using neutrons 2825 1451 154 from D(d,n)He3 reaction were renormalized to the results of this 2825 1451 155 experiment obtained with Li7(p,n)Be7 neutron source in the over- 2825 1451 156 lapping interval 5.398 - 5.870 MeV. D(d,n)He3 data in the energy 2825 1451 157 range 5.398 - 9.897 MeV were increased to the factor Fc=1.083 . 2825 1451 158 Data of Wu and Chou [15] and Husain and Hunt [20] were also 2825 1451 159 renormalized to the Smith and Meadows measurements carried out 2825 1451 160 with Li7(p,n)Be7 neutron source. Correction factors for experi- 2825 1451 161 mental data were Fc=1.04592 [15] and Fc=1.03634 [20]. 2825 1451 162 Cross sections measured by Hudson et al. [16] in the energy 2825 1451 163 range 13.3 - 17.10 MeV were corrected to Filatenkov et al. experi-2825 1451 164 mental data at 14.1 MeV [40], Fc=0.95367. 2825 1451 165 Experimental data of Rondio et al. [23,26] were corrected to 2825 1451 166 the preliminary evaluated Ni-58(n,p)Co-58m+g reaction cross sec- 2825 1451 167 tion integral in the energy interval 17.3 - 18.5 MeV (Fc=0.1889). 2825 1451 168 Data of Yuan Junqian et al. [34] and Li Tingyan et al. [35] 2825 1451 169 were multiplied to the factors Fc=0.9000 and Fc=0.9259 , respec- 2825 1451 170 tively, using preliminary evaluated cross section integral in the 2825 1451 171 energy interval 13.6 - 14.6 MeV. 2825 1451 172 Experimental cross section data [42-64] were rejected due to 2825 1451 173 their discrepancy with the main bulk of experimental data [1-41]. 2825 1451 174 Data of Decowski et al. [9] obtained in the neutron energy range 2825 1451 175 from 1.95 to 4.79 MeV were rejected for the same reason. In the 2825 1451 176 rejected experiments [42-43], [51-55], [57-60], [62-64] the cross 2825 1451 177 section values were measured only in a one energy point in the 2825 1451 178 interval 14 - 15 MeV. 2825 1451 179 Statistical analysis of input cross section data was carried 2825 1451 180 out by means of PADE-2 code [65]. Rational function was used as 2825 1451 181 the model function [66]. 2825 1451 182 U-235 thermal fission [67] and Cf-252 spontaneous fission 2825 1451 183 neutron spectra [68] averaged cross sections calculated from the 2825 1451 184 the evaluated Ni-58(n,p)Co-58m+g excitation function are the 2825 1451 185 following: 2825 1451 186 2825 1451 187 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2825 1451 188 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2825 1451 189 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2825 1451 190 U-235 neutron fission ³ 107.44 ³ 108.9 +- 5.2 [69] 2825 1451 191 ³ ³ 106.0 +- 7.0 [70] 2825 1451 192 ³ ³ 108.5 +- 1.4 [71] 2825 1451 193 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2825 1451 194 Cf-252 spont. fission ³ 117.36 ³ 117.6 +- 1.5 [72] 2825 1451 195 ³ ³ 117.5 +- 1.53 [73] 2825 1451 196 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2825 1451 197 2825 1451 198 MT=33 2825 1451 199 MT=103 -(n,p) cross section cov. matrix 2825 1451 200 --------------------------------------- 2825 1451 201 Uncertainties in the evaluated excitation function for the 2825 1451 202 reaction Ni-58(n,p)Co-58m+g are given in the form of relative co- 2825 1451 203 variance matrix for the 37-neutron energy groups (LB=5). Covari- 2825 1451 204 ance matrix of uncertainties was calculated simultaneously with 2825 1451 205 recommended cross section data by means of PADE-2 code. 2825 1451 206 Eigenvalues of the 6-th digits relative covariance matrix 2825 1451 207 given in the 33-file are the following: 2825 1451 208 2825 1451 209 3.02525E-08 3.04415E-08 3.07179E-08 3.11702E-08 2825 1451 210 3.17809E-08 3.24784E-08 3.36976E-08 3.53449E-08 2825 1451 211 3.74027E-08 4.12908E-08 4.46423E-08 5.24310E-08 2825 1451 212 5.88770E-08 7.22029E-08 8.98365E-08 1.06273E-07 2825 1451 213 1.47057E-07 1.97171E-07 2.32636E-07 3.12653E-07 2825 1451 214 4.31346E-07 5.83111E-07 7.69089E-07 9.84635E-07 2825 1451 215 1.17764E-06 1.35691E-06 6.11687E-05 6.46368E-04 2825 1451 216 7.13990E-04 1.20451E-03 1.41032E-03 1.54898E-03 2825 1451 217 2.03298E-03 3.77118E-03 4.31020E-03 5.28282E-03 2825 1451 218 3.31282E-02 2825 1451 219 2825 1451 220 References: 2825 1451 221 1. I.Kumabe, R.W.Fink Nucl. Phys., v.15, p.316, February 1960 2825 1451 222 2. R.N.Glover, E.Weigold Nucl. Phys., v.29, p.309, 1962 2825 1451 223 3. J.F.Barry J. Nucl. Energy AB, v.16, p.467, 1962 2825 1451 224 4. W.G.Cross et al. Progress Report EANDC(CAN)-16, p.1, 2825 1451 225 January 1963 2825 1451 226 5. J.W.Meadows, J.F.Whalen Phys. Rev., v.130, p.2022, 1963 2825 1451 227 6. K.Debertin, E.Rossle Nucl. Phys., v.70, p.89, August 1965 2825 1451 228 7. M.Bormann et al. Zeitschrift f. Naturforschung, section A, 2825 1451 229 v.21, p.988, 1966 ; 2825 1451 230 8. A.Paulsen, H.Liskien Proc. of the 1-st IAEA Conference on 2825 1451 231 Nuclear Data for Reactors, Paris, 17-21 October 1966, p.217 2825 1451 232 9. P.Decowski et al. Nucl. Phys. A, v.112, p.513, May 1968 2825 1451 233 10. R.C.Barrall et al. Report AFWL-TR-68-134, Albuquerque, NM, 2825 1451 234 March 1969 2825 1451 235 11. R.W.Fink, W.-D.Lu Bull. Amer. Phys. Soc., v.15, p.1372(EH6), 2825 1451 236 July 1970 2825 1451 237 12. A.Paulsen, R.Widera Proc. of Conference on Chemical Nuclear 2825 1451 238 Data, Measurements and Applicat., Univ. of Kent, Canterbury, 2825 1451 239 20-22 September 1971, p.129 2825 1451 240 13. D.L.Smith,J.W.Meadows Nucl. Sci. Eng., v.58, p.314, Nov. 1975 2825 1451 241 14. R.A.Sigg Dissertation Abstracts section B, v.37, p.2237, 2825 1451 242 November 1976 2825 1451 243 15. M.W.Wu, J.C.Chou Nucl. Sci. Eng., v.63, p.268, 1977 2825 1451 244 16. C.G.Hudson et al. Annals of Nuclear Energy, v.5, p.589, 1978 2825 1451 245 17. M.T.Swinhoe, C.A.Uttley Report AERE-R-9929, September 1980 2825 1451 246 18. P.N.Ngoc et al. Nukleonika, v.29, p.87, 1984 2825 1451 247 19. P.Raics et al. Atomki Koezlemenyek, v.23, p.45, June 1981 2825 1451 248 20. H.A.Husain, S.E.Hunt Internat. J. of Applied Radiation and 2825 1451 249 Isotopes, v.34, n.4, p.731, 1983 2825 1451 250 21. A.Pavlik et al. Nucl. Sci. Eng., v.90, p.186, June 1985 2825 1451 251 22. Lu Hanlin et al. 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Report INDC(CPR)-045, IAEA, October 1998 2825 1451 284 40. A.A.Filatenkov et al. Report RI-252, May 1999 2825 1451 285 41. T.Senga et al. Report JAERI-Conf 2000-005, p.208, March 2000 2825 1451 286 42. D.L.Allan Proc. Phys. Soc., sec.A, v.70, p.195, March 1957 2825 1451 287 43. K.H.Purser, F.W.Titterton Australian J. Physic, v.12, 2825 1451 288 p.103, 1959 2825 1451 289 44. L.Gonzalez et al. Phys. Rev., v.120, p.1319, 1960 2825 1451 290 45. K.Nakai et al. J. Phys. Soc. Japan, v.17, p.1215, July 1962 2825 1451 291 46. J.M.F.Jeronimo et al. Nucl. Phys., v.47, p.157, July 1963 2825 1451 292 47. J.Konijn, A.Lauber Nucl. Phys., v.48, p.191, November 1963 2825 1451 293 48. S.Okumura Nucl. Phys., v.A93, p.74, March 1967 2825 1451 294 49. M.Bormann et al. Proc. of 1-st IAEA Conf. on Nuclear Data for 2825 1451 295 Reactors, Paris, 17-21 October 1966, v.1, p.225, April 1967 2825 1451 296 50. J.K.Temperley Nucl. Sci. Eng., v.32, p.195, 1968 2825 1451 297 51. V.N.Levkovskij et al. 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Conf. on Nuclear 2825 1451 312 Data for Science and Technology, 30 May - 3 June 1988, Mito, 2825 1451 313 Japan, Saikon Publishing Co., LTD, pp.261-265, 1989 2825 1451 314 61. N.I.Molla et al. Proc. of Int. Conf. on Nuclear Data for 2825 1451 315 Science and Technology, Juelich, FRG, 13-17 May 1991, p. 355 2825 1451 316 62. A.Ercan et.al. Proc. of an Intern. Conf. on Nuclear Data for 2825 1451 317 Science and Technology, 13-17 May 1991, Julich, FRG, Springer-2825 1451 318 Verlag, 1992 2825 1451 319 63. I.Garlea et al. Rev. Roum. Phys., v.37, no.1, pp.19-25, 1992 2825 1451 320 64. L.I.Klochkova et al. Vop. At. Nauki i Tekhn., Ser. Yadernye 2825 1451 321 Konstanty, v.1, p.27, 1992 2825 1451 322 65. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2825 1451 323 66. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 2825 1451 324 st's Meeting on Evaluation and Processing of Covariance Data, 2825 1451 325 Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105 2825 1451 326 67. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 2825 1451 327 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 2825 1451 328 68. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2825 1451 329 69. W.Mannhart Proc. 5-th ASTM-EUR. Symp. on Reactor Dosimetry, 2825 1451 330 Geesthacht, FRG, September 24-28, 1984, Vol.2, p.813, 1985 2825 1451 331 70. O.Horibe et al. Proc. of Conf.:50 Years with Nuclear Fission, 2825 1451 332 Washington D.C., 25-28 April 1989, v.2, p.923 2825 1451 333 71. W.Mannhart Prog. Report INDC(Ger)-045, pp.40-43, June 1999 2825 1451 334 72. W.Mannhart Handbook on Nuclear Activation Cross Sections , 2825 1451 335 IAEA Technical Report Ser. No.273, p.413, 1987 2825 1451 336 73. W.Mannhart Validation of Differential Cross Sections with 2825 1451 337 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2825 1451 338 September 2002 2825 1451 339 ***************************************************************** 2825 1451 340 ******* End of RRDF (N,P) bibliographical component ******* 2825 1451 341 ***************************************************************** 2825 1451 342 The Q values and threshold energies were updated prior to pro- 2825 1451 343 cessing through the codes to comply with the values obtained 2825 1451 344 using the NNDC calculation program which is based on the 1995 2825 1451 345 Update to the Atomic mass Evaluation. 2825 1451 346 2825 1451 347 Psuedo Threshold of 1.0E+5 changed to 4.0E+5 ev in MF/MT=3/103 2825 1451 348 in original input file before processing through codes. 2825 1451 349 2825 1451 350 File 2 added to the pointwise file containing only the effective 2825 1451 351 scattering radius with no resonance parameters given. 2825 1451 352 Taken from ENDF/B-VI 2825 1451 353 ***************************************************************** 2825 1451 354 ***************** Program LINEAR (VERSION 2002-1) ***************2825 1451 355 For All Data Greater than 1.0000E-10 barns in Absolute Value 2825 1451 356 Data Linearized to Within an Accuracy of .100000000 per-cent 2825 1451 357 ***************** Program RECENT (VERSION 2002-1) ***************2825 1451 358 for All Data Greater than 1.0000E-10 barns in Absolute Value 2825 1451 359 Data Linearized to within an Accuracy of .100000000 per-cent 2825 1451 360 ***************** Program SIGMA1 (VERSION 2002-1) ***************2825 1451 361 Data Doppler Broadened to 300.000000 Kelvin 2825 1451 362 for All Data Greater than 1.0000E-10 barns in Absolute Value 2825 1451 363 Data Linearized to Within an Accuracy pf .100000000 per-cent 2825 1451 364 ***************** Program FIXUP (Version 2002-1) ****************2825 1451 365 Corrected ZA/AWR in All Sections-----------------------------Yes 2825 1451 366 Corrected Thresholds-----------------------------------------Yes 2825 1451 367 Extended Cross Sections to 20 MeV----------------------------No 2825 1451 368 Allow Cross Section Deletion---------------------------------No 2825 1451 369 Allow Cross Section Reconstruction---------------------------No 2825 1451 370 Make All Cross Sections Non-Negative-------------------------Yes 2825 1451 371 Delete Energies Not in Ascending Order-----------------------Yes 2825 1451 372 Deleted Duplicate Points-------------------------------------Yes 2825 1451 373 Check for Ascending MAT/MF/MT Order--------------------------Yes 2825 1451 374 Check for Legal MF/MT Numbers--------------------------------Yes 2825 1451 375 Allow Creation of Missing Sections---------------------------No 2825 1451 376 Allow Insertion of Energy Points-----------------------------No 2825 1451 377 Create Uniform Energy Grid-----------------------------------No 2825 1451 378 Delete Section if Cross Section =0 at All Energies-----------Yes 2825 1451 379 ***************** Program GROUPIE (VERSION 2002-1) **************2825 1451 380 Unshielded Group Averages Using 640 Groups 2825 1451 381 Weighting Spectrum: Flat (Constant) Spectrum 2825 1451 382 1 451 388 02825 1451 383 2 151 4 02825 1451 384 3 16 29 02825 1451 385 3 103 73 02825 1451 386 33 16 58 02825 1451 387 33 103 133 02825 1451 388 2825 1 099999 2825 0 0 0 2.80580E+4 5.74380E+1 0 0 1 02825 2151 1 2.805800+4 1.000000+0 0 0 1 02825 2151 2 1.000000-5 8.120000+5 0 0 0 02825 2151 3 0.000000+0 6.400000-1 0 0 0 02825 2151 4 2825 2 099999 2825 0 0 0 2.80580E+4 5.74376E+1 0 0 0 02825 3 16 1 -1.22190E+7-1.22190E+7 0 0 1 772825 3 16 2 77 1 2825 3 16 3 12400000.0 .000186234 12500000.0 .000945261 12600000.0 .0017445372825 3 16 4 12700000.0 .002543812 12800000.0 .003343087 12900000.0 .0041423622825 3 16 5 13000000.0 .005363417 13100000.0 .007006250 13200000.0 .0086490832825 3 16 6 13300000.0 .010291917 13400000.0 .011934750 13500000.0 .0135775832825 3 16 7 13600000.0 .015483500 13700000.0 .017652500 13800000.0 .0198330002825 3 16 8 13900000.0 .022025000 14000000.0 .024031250 14100000.0 .0258517502825 3 16 9 14200000.0 .028017000 14300000.0 .030527000 14400000.0 .0327212502825 3 16 10 14500000.0 .034599750 14600000.0 .036406500 14700000.0 .0381415002825 3 16 11 14800000.0 .039815500 14900000.0 .041428500 15000000.0 .0430853002825 3 16 12 15100000.0 .044785900 15200000.0 .046486500 15300000.0 .0481871002825 3 16 13 15400000.0 .049887700 15500000.0 .051421600 15600000.0 .0527888002825 3 16 14 15700000.0 .054156000 15800000.0 .055523200 15900000.0 .0568904002825 3 16 15 16000000.0 .058126600 16100000.0 .059231800 16200000.0 .0603370002825 3 16 16 16300000.0 .061442200 16400000.0 .062547400 16500000.0 .0634315002825 3 16 17 16600000.0 .064094500 16700000.0 .064757500 16800000.0 .0654205002825 3 16 18 16900000.0 .066083500 17000000.0 .066685900 17100000.0 .0672277002825 3 16 19 17200000.0 .067769500 17300000.0 .068311300 17400000.0 .0688531002825 3 16 20 17500000.0 .069477400 17600000.0 .070184200 17700000.0 .0708910002825 3 16 21 17800000.0 .071597800 17900000.0 .072304600 18000000.0 .0730607002825 3 16 22 18100000.0 .073866100 18200000.0 .074671500 18300000.0 .0754769002825 3 16 23 18400000.0 .076282300 18500000.0 .076938600 18600000.0 .0774458002825 3 16 24 18700000.0 .077953000 18800000.0 .078460200 18900000.0 .0789674002825 3 16 25 19000000.0 .079474600 19100000.0 .079981800 19200000.0 .0804890002825 3 16 26 19300000.0 .080996200 19400000.0 .081503400 19500000.0 .0820106002825 3 16 27 19600000.0 .082517800 19700000.0 .083025000 19800000.0 .0835322002825 3 16 28 19900000.0 .084039400 20000000.0 0.0 2825 3 16 29 2825 3 099999 2.80580E+4 5.74376E+1 0 0 0 02825 3103 1 4.00800E+5 4.00800E+5 0 0 1 2102825 3103 2 210 1 2825 3103 3 400000.000 9.79220E-8 425000.000 2.93766E-7 450000.000 4.89610E-72825 3103 4 475000.000 6.85454E-7 500000.000 8.96305E-7 525000.000 1.12216E-62825 3103 5 550000.000 3.02637E-6 575000.000 6.60892E-6 600000.000 1.37449E-52825 3103 6 630000.000 2.44342E-5 660000.000 3.51235E-5 690000.000 5.23130E-52825 3103 7 720000.000 9.72844E-5 760000.000 .000150537 800000.000 .0002255822825 3103 8 840000.000 .000322417 880000.000 .000426118 920000.000 .0005710122825 3103 9 960000.000 .000722772 1000000.00 .001548276 1100000.00 .0033102152825 3103 10 1200000.00 .005065710 1300000.00 .006936165 1400000.00 .0096017702825 3103 11 1500000.00 .013139050 1600000.00 .017657450 1700000.00 .0232753002825 3103 12 1800000.00 .030101100 1900000.00 .038220650 2000000.00 .0476866002825 3103 13 2100000.00 .058510800 2200000.00 .070660150 2300000.00 .0840569002825 3103 14 2400000.00 .098583350 2500000.00 .114089500 2600000.00 .1304045002825 3103 15 2700000.00 .147348500 2800000.00 .164823000 2900000.00 .1825590002825 3103 16 3000000.00 .200295000 3100000.00 .217979667 3200000.00 .2356130002825 3103 17 3300000.00 .253246333 3400000.00 .270576000 3500000.00 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4.560900-4282533103 107 4.822470-4 4.899470-4 4.712900-4 5.075800-4 4.787570-4 3.916490-4282533103 108 2.653020-4 1.347580-4 3.667320-5-5.566650-6 1.196800-5 7.748840-5282533103 109 1.717390-4 2.753360-4 3.729060-4 4.543540-4 5.144220-4 5.515150-4282533103 110 5.665130-4 5.513000-4 4.853190-4 4.324480-4 3.299840-4 2.066050-4282533103 111 9.851580-5 3.426120-5 2.566870-5 6.833550-5 1.479640-4 2.473420-4282533103 112 3.510700-4 4.476680-4 5.298930-4 5.940640-4 6.391730-4 6.707940-4282533103 113 4.268960-4 3.717220-4 2.815670-4 1.833800-4 1.050470-4 6.517590-5282533103 114 6.955450-5 1.135560-4 1.869020-4 2.778930-4 3.760410-4 4.732300-4282533103 115 5.638770-4 6.446410-4 7.413690-4 3.756610-4 3.387300-4 2.724920-4282533103 116 1.974450-4 1.345130-4 9.832300-5 9.510360-5 1.240620-4 1.800840-4282533103 117 2.562310-4 3.454750-4 4.416530-4 5.398550-4 6.807700-4 3.609990-4282533103 118 3.428820-4 2.928950-4 2.278710-4 1.658530-4 1.206730-4 1.000320-4282533103 119 1.060900-4 1.371230-4 1.892260-4 2.576610-4 3.377130-4 4.689640-4282533103 120 3.773120-4 3.706070-4 3.302750-4 2.703650-4 2.058530-4 1.486660-4282533103 121 1.063320-4 8.229690-5 7.695740-5 8.879390-5 1.152710-4 1.761240-4282533103 122 4.157570-4 4.230810-4 3.969360-4 3.463240-4 2.815830-4 2.118020-4282533103 123 1.437810-4 8.194610-5 2.873500-5-1.490280-5-6.107470-5 4.907200-4282533103 124 5.238500-4 5.211570-4 4.859640-4 4.245560-4 3.440740-4 2.512660-4282533103 125 1.518550-4 5.033710-5-9.636020-5 6.307180-4 7.025830-4 7.333670-4282533103 126 7.226360-4 6.745020-4 5.953670-4 4.923350-4 3.722470-4 1.753660-4282533103 127 8.655530-4 9.914960-4 1.072120-3 1.105210-3 1.093810-3 1.043780-3282533103 128 9.621740-4 7.967120-4 1.233200-3 1.438380-3 1.597930-3 1.708340-3282533103 129 1.771340-3 1.791690-3 1.753320-3 1.793410-3 2.115770-3 2.395310-3282533103 130 2.626540-3 2.809320-3 2.992600-3 2.629110-3 3.114210-3 3.559090-3282533103 131 3.955790-3 4.443820-3 3.832780-3 4.524150-3 5.174450-3 6.033490-3282533103 132 5.486660-3 6.417000-3 7.695010-3 7.646890-3 9.373270-3 1.179240-2282533103 133 282533 099999 2825 0 0 0 0 0 0 0 2.80600E+4 5.94159E+1 0 0 34 102831 1451 1 0.0 0.0 0 0 0 62831 1451 2 1.00000E+0 1.50000E+8 0 0 10 20022831 1451 3 3.00000E+2 0.0 1 0 395 42831 1451 4 28-Ni- 60 LANL,ORNL EVAL-SEP97 S.CHIBA,M.B.CHADWICK,LARSON 2831 1451 5 DIST-Feb2004 2831 1451 6 ----IRDF-2002 MATERIAL 2831 2831 1451 7 -----INCIDENT NEUTRON DATA 2831 1451 8 ------ENDF-6 FORMAT 2831 1451 9 2831 1451 10 ******************************************************************2831 1451 11 ----ENDF/B-VI MATERIAL 2831 REVISION 3 2831 1451 12 Ch97,Ch99 DIST-SEP 1 REV3- 20010926 2831 1451 13 2831 1451 14 ENDF/B-VI MOD 4 Revision, June 2000, S.C. Frankle, R.C. Reedy, 2831 1451 15 P.G. Young (LANL) 2831 1451 16 2831 1451 17 The secondary gamma-ray spectrum for radiative capture (MF 12, 2831 1451 18 MT 102) has been updated for new experimental data at incident 2831 1451 19 neutron energies up to 1 keV. 2831 1451 20 The Q-value for radiative capture was also updated in File 3. 2831 1451 21 Details of these changes are described in Frankel et al. [Fr01]. 2831 1451 22 2831 1451 23 ******************************************************************2831 1451 24 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2831 1451 25 DOSIMETRY 2831 1451 26 ******************************************************************2831 1451 27 2831 1451 28 ENDF/B-VI MOD 3 Evaluation, September 1997, S. Chiba, 2831 1451 29 M.B. Chadwick, P.G. Young (LANL), and 2831 1451 30 A.J. Koning (ECN) 2831 1451 31 2831 1451 32 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code 2831 1451 33 in cooperation with ECN Petten. 2831 1451 34 2831 1451 35 This evaluation provides a complete representation of the 2831 1451 36 nuclear data needed for transport, damage, heating, 2831 1451 37 radioactivity, and shielding applications over the incident 2831 1451 38 neutron energy range from 1.0E-11 to 150 MeV. The discussion 2831 1451 39 here is divided into the region below and above 20 MeV. 2831 1451 40 2831 1451 41 INCIDENT NEUTRON ENERGIES < 20 MeV 2831 1451 42 2831 1451 43 Below 20 MeV the evaluation is based completely on the ENDF/B- 2831 1451 44 VI.5(Rev1,Release 2) evaluation by Larson, C. Perey, Hetrich, and 2831 1451 45 Fu. 2831 1451 46 2831 1451 47 INCIDENT NEUTRON ENERGIES > 20 MeV 2831 1451 48 2831 1451 49 The ENDF/B-VI Release 2 evaluation extends to 20 MeV and 2831 1451 50 includes cross sections and energy-angle data for all 2831 1451 51 significant reactions. The present evaluation utilizes a more 2831 1451 52 compact composite reaction spectrum representation above 20 MeV 2831 1451 53 in order to reduce the length of the file. No essential data for 2831 1451 54 applications is lost with this representation. 2831 1451 55 The evaluation above 20 MeV utilizes MF=6, MT=5 to represent 2831 1451 56 all reaction data. Production cross sections and emission 2831 1451 57 spectra are given for neutrons, protons, deuterons, tritons, 2831 1451 58 alpha particles, gamma rays, and all residual nuclides produced 2831 1451 59 (A>5) in the reaction chains. To summarize, the ENDF sections 2831 1451 60 with non-zero data above En = 20 MeV are: 2831 1451 61 2831 1451 62 MF=3 MT= 1 Total Cross Section 2831 1451 63 MT= 2 Elastic Scattering Cross Section 2831 1451 64 MT= 3 Nonelastic Cross Section 2831 1451 65 MT= 5 Sum of Binary (n,n') and (n,x) Reactions 2831 1451 66 2831 1451 67 MF=4 MT= 2 Elastic Angular Distributions 2831 1451 68 2831 1451 69 MF=6 MT= 5 Production Cross Sections and Energy-Angle 2831 1451 70 Distributions for Emission Neutrons, Protons, 2831 1451 71 Deuterons, Tritons, and Alphas; and Angle- 2831 1451 72 Integrated Spectra for Gamma Rays and Residual 2831 1451 73 Nuclei That Are Stable Against Particle Emission 2831 1451 74 2831 1451 75 The evaluation is based on nuclear model calculations that 2831 1451 76 have been benchmarked to experimental data, especially for n + 2831 1451 77 Ni58 and p + Ni58 reactions [Ch97]. We use the GNASH code system 2831 1451 78 [Yo92], which utilizes Hauser-Feshbach statistical, 2831 1451 79 preequilibrium and direct-reaction theories. Spherical optical 2831 1451 80 model calculations are used to obtain particle transmission 2831 1451 81 coefficients for the Hauser-Feshbach calculations, as well as 2831 1451 82 for the elastic neutron angular distributions. 2831 1451 83 Cross sections and spectra for producing individual residual 2831 1451 84 nuclei are included for reactions. The energy-angle-correlations 2831 1451 85 for all outgoing particles are based on Kalbach systematics 2831 1451 86 [Ka88]. 2831 1451 87 A model was developed to calculate the energy distributions of 2831 1451 88 all recoil nuclei in the GNASH calculations [Ch96]. The recoil 2831 1451 89 energy distributions are represented in the laboratory system in 2831 1451 90 MT=5, MF=6, and are given as isotropic in the lab system. All 2831 1451 91 other data in MT=5,MF=6 are given in the center-of-mass system. 2831 1451 92 This method of representation utilizes the LCT=3 option approved 2831 1451 93 at the November, 1996, CSEWG meeting. 2831 1451 94 Preequilibrium corrections were performed in the course of the 2831 1451 95 GNASH calculations using the exciton model of Kalbach [Ka77, 2831 1451 96 Ka85], validated by comparison with calculations using Feshbach, 2831 1451 97 Kerman, Koonin (FKK) theory [Ch93]. Discrete level data from 2831 1451 98 nuclear data sheets were matched to continuum level densities 2831 1451 99 using the formulation of Gilbert and Cameron [Gi65] and pairing 2831 1451 100 and shell parameters from the Cook [Co67] analysis. Neutron and 2831 1451 101 charged- particle transmission coefficients were obtained from 2831 1451 102 the optical potentials, as discussed below. Gamma-ray 2831 1451 103 transmission coefficients were calculated using the Kopecky-Uhl 2831 1451 104 model [Ko90]. 2831 1451 105 2831 1451 106 SOME Ni-SPECIFIC INFORMATION CONCERNING THE EVAL. 2831 1451 107 2831 1451 108 The neutron total cross section was evaluated based on the 2831 1451 109 least-squares method taking account of the experimental 2831 1451 110 data [Du67, Bo71, St71, Sm79, Fe80, Pe82, Ha82a, Di97]. The data 2831 1451 111 for natural Ni [Di97] was also used because there was not enough 2831 1451 112 data for Ni-60 above 20 MeV. The data for natural Ni were 2831 1451 113 transformed to the Ni-60 cross section according to A*(2/3) law. 2831 1451 114 Result of this estimation was used as the evaluated total cross 2831 1451 115 section data above 20 MeV. 2831 1451 116 2831 1451 117 The evaluated total cross section data (1 to 250 MeV), s-wave 2831 1451 118 strength function [Mu81] and elastic scattering angular 2831 1451 119 distribution data [Bo71, Gu85, Tu73, Ya79] were used to obtain the2831 1451 120 neutron optical potential parameters. The parameter estimation was2831 1451 121 carried out based on Marquart-Bayesian approach [Sm91], where 2831 1451 122 ECIS95 code [Ra96] was used for the optical model calculation. We2831 1451 123 have employed the energy dependence of the optical potential 2831 1451 124 similar to Delaroche's work [De89]. The initial potential 2831 1451 125 parameters were adopted from Koning and Delaroche [Ko97]. A total2831 1451 126 of 7 parameters concerning the central potential depth were 2831 1451 127 estimated with associated covariance matrix, while the geometrical2831 1451 128 parameters were fixed to the result of a similar search for n + 2831 1451 129 Ni-58. Presently obtained potential was used for the calculation 2831 1451 130 of neutron transmission coefficients and DWBA cross sections in 2831 1451 131 the energy region above 20 MeV. Below 20 MeV, the Harper neutron 2831 1451 132 potential [Ha82b] was used for the calculation of transmission 2831 1451 133 coefficients. 2831 1451 134 2831 1451 135 The proton optical potential was also searched to obtain a good 2831 1451 136 description of proton-total reaction cross section as predicted by2831 1451 137 Wellisch-Axen systematic [We96] above 50 MeV. The parameter 2831 1451 138 estimation was carried out by the Marquart-Bayesian approach 2831 1451 139 similar to the neutron OMP, but trying to seek the best parameter 2831 1451 140 to reproduce the reaction cross sections compiled by Carlson 2831 1451 141 [Ca96] and Wellisch values. The experimental data in Carlson 2831 1451 142 [Ca96] was scaled for Ni-60 according to A**(2/3) law. In this 2831 1451 143 search, the geometrical parameters were fixed to be same as the 2831 1451 144 neutron potential. The present potential gives a good description2831 1451 145 of the proton total reaction cross section from 10 MeV to 250 MeV.2831 1451 146 However, after some trial and error to reproduce both the elastic 2831 1451 147 scattering and reaction cross section data for Ni-58, we have 2831 1451 148 employed the following combination of proton potentials: 2831 1451 149 2831 1451 150 0 to 5 MeV : Harper potential [Ha82b] 2831 1451 151 6 to 47 MeV : Koning and Delaroche [Ko97] 2831 1451 152 48 to 260 MeV : Present OMP 2831 1451 153 2831 1451 154 For deuterons, the Lohr-Haeberli [Lo74] global potential was used;2831 1451 155 for alpha particles the McFadden-Satchler [Mc66] potential was 2831 1451 156 used; and for tritons the Becchetti-Greenlees [Be71] potential was2831 1451 157 used. The He-3 channel was ignored. 2831 1451 158 2831 1451 159 The direct collective inelastic scattering to the following level 2831 1451 160 in Ni-60 was considered by the DWBA-mode calculation of ECIS95 2831 1451 161 [Ra96]: 2831 1451 162 2831 1451 163 Jpi Ex(MeV) Deformation length 2831 1451 164 2+ 1.331 0.8535 2831 1451 165 2831 1451 166 The deformation length was determined to much the ENDF/B-VI value 2831 1451 167 at 20 MeV. 2831 1451 168 2831 1451 169 Certain nuclear level densities were modified from the default 2831 1451 170 values in accordance with experimental level-density information 2831 1451 171 as follows. The level density of 57Fe was matched to the observed 2831 1451 172 D-0 value by modifying the pairing energy to 0.15 MeV; the level 2831 1451 173 density of the main competition channel, neutron emission to 60Ni,2831 1451 174 was adjusted to match Fischer et al.'s [Fi86] value of 3.85e3+-25%2831 1451 175 at 11 MeV excitation energy, but then increased slightly (a total 2831 1451 176 density increase of 11%) by using a pairing energy of 1.1 MeV, 2831 1451 177 since Haight's alpha production at 14 MeV is approx. 11% smaller 2831 1451 178 than Fischer et al.'s measurements. Production of residual nuclei 2831 1451 179 59Ni and 56Fe, through (n,2n) and (n,na) reactions, become 2831 1451 180 particularly important above 14 MeV. Pairing energies were 2831 1451 181 adjusted to match the level density for 59Ni at the neutron 2831 1451 182 binding energy (D0=12.9 keV), and Fischer et al.'s result for the 2831 1451 183 56Fe level density at 11 MeV. 2831 1451 184 2831 1451 185 The new Haight et al. [Ha97] LANL/WNR 60Ni(n,x alpha) data up to 2831 1451 186 50 MeV was used to benchmark our model calculations [Ch97] - 2831 1451 187 agreement with experiment was good. 2831 1451 188 2831 1451 189 **************************************************************** 2831 1451 190 2831 1451 191 REFERENCES 2831 1451 192 2831 1451 193 [Be71] F.D. Becchetti, Jr., and G.W. Greenlees in 2831 1451 194 "Polarization Phenomena in Nuclear Reactions," (Ed: H.H. 2831 1451 195 Barschall and W. Haeberli, The University of Wisconsin Press, 2831 1451 196 1971) p.682 2831 1451 197 [Bo71] P. Boschung et al, Nucl.Phys. A161, 593 (1971) 2831 1451 198 [Ca96] R.F. Carlson, Atomic Data and Nuclear Data Tables, 63, 2831 1451 199 93 (1996) 2831 1451 200 [Ch93] M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 2831 1451 201 [Ch96] M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J. 2831 1451 202 Koning, "High-Energy Nuclear Data Libraries for Accelerator- 2831 1451 203 Driven Technologies: Calculational Method for Heavy Recoils," 2831 1451 204 Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation 2831 1451 205 Technology and Applications, Kalmar, Sweden, 3-7 June 1996 2831 1451 206 [Ch97] M.B. Chadwick and P.G. Young, "Model Calculations of 2831 1451 207 n,p + 58,60,61,62,64Ni" in APT PROGRESS REPORT: 1 August - 1 2831 1451 208 September 1997, internal Los Alamos National Laboratory memo T-2831 1451 209 2-97/MS-51, 8 September 1997 from R.E. MacFarlane to L. Waters.2831 1451 210 [Ch99] M.B. Chadwick, P.G. Young, G.M. Hale, et al., Los Alamos 2831 1451 211 National Laboratory report, LA-UR-99-1222 (1999) 2831 1451 212 [Ci68] S. Cierjack et al, report KFK-1000 (1968) 2831 1451 213 [Co67] J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J. 2831 1451 214 Phys. 20, 477 (1967) 2831 1451 215 [De89] J.P. Delaroche, Y. Wang and J. Rapaport, Phys.Rev.C 39, 2831 1451 216 391 (1989) 2831 1451 217 [Di97] F. Dietrich et al., private communication (1997). 2831 1451 218 [Du67] Yu.V. Dukarevich et al., Nucl.Phys. A92, 433 (1967) 2831 1451 219 [Fe80] M.B. Fedorov et al., 80Kiev, 1, 309(1980) 2831 1451 220 [Fi86] R. Fischer, G. Traxler, M. Uhl at al., Phys Rev C34, 460 2831 1451 221 (1986) 2831 1451 222 [Fr01] S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos 2831 1451 223 National Laboratory Report, LA-13812 (2001). 2831 1451 224 [Fu96] C.Y. Fu, "Effects of shape differences in the level 2831 1451 225 densities of three formalisms on calculated cross sections", 2831 1451 226 personal communication, 1996 (to be published by the IAEA). 2831 1451 227 [Gi65] A. Gilbert and A.G.W. Cameron, Can.J.Phys. 43, 1446 (1965)2831 1451 228 [Gu85] P.P. Guss et al, Nucl.Phys. A438, 187 (1985) 2831 1451 229 [Ha82a] J.A. Harvey et al, 82Anterp (1982) p. 856 2831 1451 230 [Ha82b] R.C. Harper and W.L. Alford, J.Phys.G. 8, 153 (1982) 2831 1451 231 [Ha97] R.C. Haight, F.B. Bateman, S.M. Sterbenz et al., Nuclear 2831 1451 232 Data for Science and Technol., Proc. Conf. Trieste, Italy, May 2831 1451 233 1997, Vol. I (Edit. Compositori, Bologna, Italy, 1997) p. 603 2831 1451 234 [Ka77] C. Kalbach, Z.Phys.A 283, 401 (1977) 2831 1451 235 [Ka85] C. Kalbach, Los Alamos National Laboratory report 2831 1451 236 LA-10248-MS (1985) 2831 1451 237 [Ka88] C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also 2831 1451 238 C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981) 2831 1451 239 [Ko90] J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990) 2831 1451 240 [Ko97] A. Koning and J.P. Delaroche, private communication. 2831 1451 241 [La83] D.C. Larson et al, report ORNL-TM-8203 (1983) 2831 1451 242 [Lo74] J.M.Lohr and W.Haeberli, Nucl.Phys. A232, 381 (1974) 2831 1451 243 [Mc66] L. McFadden and G.R. Satchler, Nucl.Phys. 84, 177 (1966) 2831 1451 244 [Mu81] S.F. Mughabghab, M. Divadeenam and N.E. Holden, "Neutron 2831 1451 245 Cross Sections", Vol. 1, Part A (Academic Press, 1981) 2831 1451 246 [Pe73] F.G. Perey, private communication (1973) [EXFOR 10342] 2831 1451 247 [Pe82] C.M. Perey et al, report ORNL-5893 (1982) 2831 1451 248 [Pe88] Pedroni et al, Phys.Rev.C 38, 2052 (1988) 2831 1451 249 [Po81] W. Poenitz, Proc. Conf. on Nuclear Data Evaluation Methods2831 1451 250 and Procedures, Brookhaven National Laboratory Report BNL-NCS- 2831 1451 251 51363, p.249(1981). 2831 1451 252 [Ra96] J. Raynal, "Notes on ECIS94", Service de Physique 2831 1451 253 Theorique, Saclay, France (personal communication through A. J.2831 1451 254 Koning, 1996). 2831 1451 255 [Sc73] W. Schimmerling et al., Phys.Rev.C 7, 248 (1973) 2831 1451 256 [Sm79] A.B. Smith et al., Nucl.Sci.Eng., 72, 293 (1979) 2831 1451 257 [Sm91] D.L. Smith, "Probability, Statistics, and Data Uncertainty2831 1451 258 in Nuclear Science and Technology" (American Nuclear Society, 2831 1451 259 1991) 2831 1451 260 [Sm92] A.B. Smith et al., J.Phys.G, 18, 629 (1992) 2831 1451 261 [St71] P. Stoler et al,, 71Knox, 1, 311 (1971) 2831 1451 262 [Tu73] A.I. Tutubalin et al., Neutron Physics, Proc. Conf., Kiev,2831 1451 263 Ukraine, 1973, Vol. 3 [EXFOR 40417] p. 62 2831 1451 264 [Ul83] J.J. Ullmann, F.P. Brady, C.M. Castaneda, et al., Nucl. 2831 1451 265 Phys. A427, 493 (1984) 2831 1451 266 [We96] H.P. Wellisch and D. Axen, Phys.Rev.C 54, 1329 (1996) 2831 1451 267 [Ya79] Y. Yamanouti, J. Rapaport, S.M. Grimes et al, Nuclear 2831 1451 268 Cross Sections for Technology, Proc. Conf., Knoxville, TN, 2831 1451 269 1979, NBS Special Publication 594 (1980) [EXFOR 10953] 2831 1451 270 [Yo92] P.G. Young, E.D. Arthur, and M.B. Chadwick, Los Alamos 2831 1451 271 report LA-12343-MS (1992) 2831 1451 272 2831 1451 273 **************************************************************** 2831 1451 274 2831 1451 275 ENDF/B-VI MOD 2 Revision, July 1991, D.C. Larson, C.M. Perey, 2831 1451 276 D.M. Hetrick, and C.Y. Fu (ORNL) 2831 1451 277 2831 1451 278 REVISION 1 CHANGES 2831 1451 279 The resonance region from 1.E-5 eV to 450 keV has been changed,2831 1451 280 based on a SAMMY analysis by F. G. Perey of new 60Ni ORELA data 2831 1451 281 up to 100 keV of Harvey, Hill and Perey. This new work reproduces2831 1451 282 the thermal cross sections and includes the small resonances 2831 1451 283 inadvertently left out in the original file. No background 2831 1451 284 cross sections are required in file 3 in the resonance region. 2831 1451 285 Removed elastic transformation matrix. 2831 1451 286 2831 1451 287 **************************************************************** 2831 1451 288 2831 1451 289 ENDF/B-VI MOD 1 Evaluation, October 1989, D.C. Larson, 2831 1451 290 C.M. Perey, D.M. Hetrick, and C.Y. Fu (ORNL) 2831 1451 291 2831 1451 292 This work employed nuclear model codes including the 2831 1451 293 Distorted Wave Born Approximation (DWBA) program DWUCK (1) 2831 1451 294 and the Hauser-Feshbach code TNG (2,3,4). The TNG code provides 2831 1451 295 energy and angular distributions of particles emitted in the 2831 1451 296 compound and pre-compound reactions, ensures consistency among 2831 1451 297 all reactions, and maintains energy balance. Details pertinent 2831 1451 298 to the contents of this evaluation and extensive comparisons of 2831 1451 299 calculations with experimental data can be found in Hetrick (5). 2831 1451 300 2831 1451 301 ----- DESCRIPTION OF FILES 2831 1451 302 (MF-MT) 2831 1451 303 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2831 1451 304 2-151 RESONANCE PARAMETERS -- TAKEN FROM REF (6). THE NEGATIVE 2831 1451 305 ENERGY RESONANCES WERE ADJUSTED BY THE AUTHOR, C. M. PEREY2831 1451 306 TO GIVE THE PROPER THERMAL SCATTERING AND CAPTURE. USING 2831 1451 307 THESE MODIFIED RESULTS, NO BACKGROUND IS REQUIRED IN 3/1 2831 1451 308 IE, THE TOTAL, SCATTERING AND CAPTURE IS GIVEN COMPLETELY 2831 1451 309 BY THE RESONANCE PARAMETERS FROM 1.E-5 EV TO 450 KEV. 2831 1451 310 RESONANCE PARAMETERS ARE FROM A REICH-MOORE ANALYSIS WITH 2831 1451 311 THE CODE SAMMY (REF 7). 2831 1451 312 THUS, THE THERMAL CROSS SECTIONS ARE GIVEN BY THE 2831 1451 313 RESONANCE PARAMETERS AND HAVE VALUES: TOTAL 3.9 B, 2831 1451 314 ELASTIC SCATTERING 0.98 B, AND CAPTURE 2.92 B. 2831 1451 315 NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2831 1451 316 OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE 2831 1451 317 PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON 2831 1451 318 A FINER ENERGY GRID THAN GIVEN IN 4/2. 2831 1451 319 3-103 (N,P) CROSS SECTIONS FOR EN LESS THAN OR EQUAL TO 12 MEV 2831 1451 320 WERE TAKEN FROM A CURVE DRAWN THROUGH THE DATA OF VONACH 2831 1451 321 ET AL. (10). THIS CURVE WAS CONNECTED TO THE ANL DATA OF 2831 1451 322 GREENWOOD (11) FROM 14.5 TO 15.0 MEV. FROM 15.0 TO 20.0 2831 1451 323 MEV THE DATA OF PAULSEN AND LISKIEN (12) WERE USED 2831 1451 324 AS A GUIDE. THE NEW DATA BY VONACH ET AL. (10) 2831 1451 325 RESULT IN AN INTEGRAL VALUE CLOSE TO THE 2.39 VALUE OF 2831 1451 326 MANNHART (13). 2831 1451 327 -------------------------------------------------------------- 2831 1451 328 UNCERTAINTY FILES 2831 1451 329 AN LB=8 SECTION IS INCLUDED FOR ALL NON-DERIVED FILES AS 2831 1451 330 REQUIRED BY ENDF/B-VI. 2831 1451 331 2831 1451 332 33-103 COVARIANCES - DATA OF VONACH ET AL. (10) AND CSISRS DATA 2831 1451 333 USED AS A GUIDE. 2831 1451 334 ---------------------------------------------------------------- 2831 1451 335 REFERENCES: 2831 1451 336 (1) P.D. Kunz, "DISTORTED WAVE CODE DWUCK72," UNIV. OF 2831 1451 337 COLORADO, UNPUBLISHED (1972). 2831 1451 338 (2) C.Y. Fu, Oak Ridge report ORNL/TM-7042 (1980) 2831 1451 339 (3) C.Y Fu, Neutron Cross Sections from 10 to 50 MeV, Proc. Symp.2831 1451 340 Upton, NY, May 1980, Brookhaven report BNL-NCS-51425 (1980) 2831 1451 341 p. 675 2831 1451 342 (4) K. Shibata and C.Y. Fu, Oak Ridge National Laboratory report 2831 1451 343 ORNL/TM-10093 (1986) 2831 1451 344 (5) D.M. Hetrick, C.Y. Fu, and D.C. Larson, Oak Ridge National 2831 1451 345 Laboratory report ORNL/TM-10219 [ENDF-344] (1987) 2831 1451 346 (6) C.M. Perey, J.A. Harvey, R.L. Macklin, et al., Phys.Rev.C, 2831 1451 347 27, 2556 (1983) 2831 1451 348 (7) N.M. Larson and F.G. Perey, Oak Ridge National Laboratory 2831 1451 349 reports ORNL/TM-7485 (1980); ORNL/TM-9179 (1984); 2831 1451 350 ORNL/TM-9179/R1 (1985); ORNL/TM-9179/R2 (1988) 2831 1451 351 (8) J.A. Harvey, ORELA, 80-M NE110 TA TARGET DATA, SENT TO NNDC 2831 1451 352 (9) J.A. Harvey, ORELA, 80-M NE110 BE BLOCK DATA, SENT TO NNDC 2831 1451 353 (10) H. Vonach, M. Wagner, and R.C. Haight, "Neutron cross 2831 1451 354 sections of 58Ni and 60Ni for 8-12 MeV neutrons," private 2831 1451 355 communication, 1989 2831 1451 356 (11) L.R. Greenwood, report DOE-ER-0046-21 (1985) p. 15 2831 1451 357 (12) A. Paulsen, Nukleonik 10, 91 (1967) 2831 1451 358 (13) W. Mannhart, Proc. Fourth ASTM-Euratom Symp. on Reactor 2831 1451 359 Dosimetry, Vol. II, March 22-26, 1982 (1982) p. 637 2831 1451 360 (14) M. Divadeenam, Brookhaven National Laboratory report 2831 1451 361 BNL-NCS-51346 [ENDF-294] (1979) 2831 1451 362 (15) S.M. Grimes, R.C. Haight, K.R. Alvar, et al., Phys.Rev.C 2831 1451 363 19, 2127 (1979) 2831 1451 364 ****************************************************************2831 1451 365 Psuedo Threshold of 0.0 added at 2.7E+6 ev to MF/MT=3/103 after 2831 1451 366 processing thro. codes. 2831 1451 367 2831 1451 368 2831 1451 369 2831 1451 370 ************************ C O N T E N T S *********************** 2831 1451 371 ***************** Program LINEAR (VERSION 2002-1) ***************2831 1451 372 For All Data Greater than 1.0000E-10 barns in Absolute Value 2831 1451 373 Data Linearized to Within an Accuracy of .100000000 per-cent 2831 1451 374 ***************** Program RECENT (VERSION 2002-1) ***************2831 1451 375 for All Data Greater than 1.0000E-10 barns in Absolute Value 2831 1451 376 Data Linearized to within an Accuracy of .100000000 per-cent 2831 1451 377 ***************** Program SIGMA1 (VERSION 2002-1) ***************2831 1451 378 Data Doppler Broadened to 300.000000 Kelvin 2831 1451 379 for All Data Greater than 1.0000E-10 barns in Absolute Value 2831 1451 380 Data Linearized to Within an Accuracy pf .100000000 per-cent 2831 1451 381 ***************** Program FIXUP (Version 2002-1) ****************2831 1451 382 Corrected ZA/AWR in All Sections-----------------------------Yes 2831 1451 383 Corrected Thresholds-----------------------------------------Yes 2831 1451 384 Extended Cross Sections to 20 MeV----------------------------No 2831 1451 385 Allow Cross Section Deletion---------------------------------No 2831 1451 386 Allow Cross Section Reconstruction---------------------------No 2831 1451 387 Make All Cross Sections Non-Negative-------------------------Yes 2831 1451 388 Delete Energies Not in Ascending Order-----------------------Yes 2831 1451 389 Deleted Duplicate Points-------------------------------------Yes 2831 1451 390 Check for Ascending MAT/MF/MT Order--------------------------Yes 2831 1451 391 Check for Legal MF/MT Numbers--------------------------------Yes 2831 1451 392 Allow Creation of Missing Sections---------------------------No 2831 1451 393 Allow Insertion of Energy Points-----------------------------No 2831 1451 394 Create Uniform Energy Grid-----------------------------------No 2831 1451 395 Delete Section if Cross Section =0 at All Energies-----------Yes 2831 1451 396 ***************** Program GROUPIE (VERSION 2002-1) **************2831 1451 397 Unshielded Group Averages Using 640 Groups 2831 1451 398 Weighting Spectrum: Flat (Constant) Spectrum 2831 1451 399 1 451 403 42831 1451 400 2 151 4 42831 1451 401 3 103 61 12831 1451 402 33 103 14 12831 1451 403 2831 1 099999 2831 0 0 0 2.80600E+4 5.94160E+1 0 0 1 02831 2151 1 2.806000+4 1.000000+0 0 0 1 02831 2151 2 1.000000-5 4.500000+5 0 0 0 02831 2151 3 0.000000+0 6.000000-1 0 0 0 02831 2151 4 2831 2 099999 2831 0 0 0 2.80600E+4 5.94159E+1 0 0 0 02831 3103 1 -2.04150E+6-2.04150E+6 0 0 1 1742831 3103 2 174 1 2831 3103 3 2700000.00 1.27290E-8 2800000.00 3.81870E-8 2900000.00 6.36450E-82831 3103 4 3000000.00 2.39704E-6 3100000.00 7.03836E-6 3200000.00 1.16797E-52831 3103 5 3300000.00 1.63210E-5 3400000.00 2.09623E-5 3500000.00 5.81607E-52831 3103 6 3600000.00 .000127916 3700000.00 .000197672 3800000.00 .0002674272831 3103 7 3900000.00 .000337182 4000000.00 .000553104 4100000.00 .0009151922831 3103 8 4200000.00 .001277280 4300000.00 .001639368 4400000.00 .0020014562831 3103 9 4500000.00 .002751840 4600000.00 .003890520 4700000.00 .0050292002831 3103 10 4800000.00 .006167880 4900000.00 .007306560 5000000.00 .0085883102831 3103 11 5100000.00 .010013130 5200000.00 .011437950 5300000.00 .0128627702831 3103 12 5400000.00 .014287590 5500000.00 .015850000 5600000.00 .0175500002831 3103 13 5700000.00 .019250000 5800000.00 .020950000 5900000.00 .0226500002831 3103 14 6000000.00 .024550000 6100000.00 .026650000 6200000.00 .0287500002831 3103 15 6300000.00 .030850000 6400000.00 .032950000 6500000.00 .0350000002831 3103 16 6600000.00 .037000000 6700000.00 .039000000 6800000.00 .0410000002831 3103 17 6900000.00 .043000000 7000000.00 .045200000 7100000.00 .0476000002831 3103 18 7200000.00 .050000000 7300000.00 .052400000 7400000.00 .0548000002831 3103 19 7500000.00 .056900000 7600000.00 .058700000 7700000.00 .0605000002831 3103 20 7800000.00 .062300000 7900000.00 .064100000 8000000.00 .0660000002831 3103 21 8100000.00 .068000000 8200000.00 .070000000 8300000.00 .0720000002831 3103 22 8400000.00 .074000000 8500000.00 .075800000 8600000.00 .0774000002831 3103 23 8700000.00 .079000000 8800000.00 .080600000 8900000.00 .0822000002831 3103 24 9000000.00 .083950000 9100000.00 .085850000 9200000.00 .0877500002831 3103 25 9300000.00 .089650000 9400000.00 .091550000 9500000.00 .0933500002831 3103 26 9600000.00 .095050000 9700000.00 .096750000 9800000.00 .0984500002831 3103 27 9900000.00 .100150000 10000000.0 .101850000 10100000.0 .1035500002831 3103 28 10200000.0 .105250000 10300000.0 .106950000 10400000.0 .1086500002831 3103 29 10500000.0 .110350000 10600000.0 .112050000 10700000.0 .1137500002831 3103 30 10800000.0 .115450000 10900000.0 .117150000 11000000.0 .1188500002831 3103 31 11100000.0 .120550000 11200000.0 .122250000 11300000.0 .1239500002831 3103 32 11400000.0 .125650000 11500000.0 .127350000 11600000.0 .1290500002831 3103 33 11700000.0 .130750000 11800000.0 .132450000 11900000.0 .1341500002831 3103 34 12000000.0 .135550000 12100000.0 .136650000 12200000.0 .1377500002831 3103 35 12300000.0 .138850000 12400000.0 .139950000 12500000.0 .1410500002831 3103 36 12600000.0 .142150000 12700000.0 .143250000 12800000.0 .1443500002831 3103 37 12900000.0 .145450000 13000000.0 .146200000 13100000.0 .1466000002831 3103 38 13200000.0 .147000000 13300000.0 .147400000 13400000.0 .1478000002831 3103 39 13500000.0 .147850000 13600000.0 .147550000 13700000.0 .1472500002831 3103 40 13800000.0 .146950000 13900000.0 .146650000 14000000.0 .1460520002831 3103 41 14100000.0 .145156000 14200000.0 .144260000 14300000.0 .1433640002831 3103 42 14400000.0 .142468000 14500000.0 .140559000 14600000.0 .1376370002831 3103 43 14700000.0 .134715000 14800000.0 .131793000 14900000.0 .1288710002831 3103 44 15000000.0 .125949000 15100000.0 .123027000 15200000.0 .1201050002831 3103 45 15300000.0 .117183000 15400000.0 .114261000 15500000.0 .1113390002831 3103 46 15600000.0 .108417000 15700000.0 .105495000 15800000.0 .1025730002831 3103 47 15900000.0 .099651000 16000000.0 .097483667 16100000.0 .0960710002831 3103 48 16200000.0 .094658333 16300000.0 .093245667 16400000.0 .0918330002831 3103 49 16500000.0 .090420333 16600000.0 .089007667 16700000.0 .0875950002831 3103 50 16800000.0 .086182333 16900000.0 .084769667 17000000.0 .0833570002831 3103 51 17100000.0 .081944333 17200000.0 .080531667 17300000.0 .0791190002831 3103 52 17400000.0 .077706333 17500000.0 .076620000 17600000.0 .0758600002831 3103 53 17700000.0 .075100000 17800000.0 .074340000 17900000.0 .0735800002831 3103 54 18000000.0 .072820000 18100000.0 .072060000 18200000.0 .0713000002831 3103 55 18300000.0 .070540000 18400000.0 .069780000 18500000.0 .0690200002831 3103 56 18600000.0 .068260000 18700000.0 .067500000 18800000.0 .0667400002831 3103 57 18900000.0 .065980000 19000000.0 .065220000 19100000.0 .0644600002831 3103 58 19200000.0 .063700000 19300000.0 .062940000 19400000.0 .0621800002831 3103 59 19500000.0 .061420000 19600000.0 .060660000 19700000.0 .0599000002831 3103 60 19800000.0 .059140000 19900000.0 .058380000 20000000.0 0.0 2831 3103 61 2831 3 099999 2831 0 0 0 2.80600E+4 5.94160E+1 0 0 0 1283133103 1 0.000000+0 0.000000+0 0 103 0 4283133103 2 0.000000+0 0.000000+0 0 0 6 3283133103 3 1.000000-5 0.000000+0 2.700000+6 1.00000-11 2.000000+7 0.000000+0283133103 4 0.000000+0 0.000000+0 0 1 6 3283133103 5 1.000000-5 0.000000+0 2.700000+6 5.000000-3 2.000000+7 0.000000+0283133103 6 0.000000+0 0.000000+0 0 1 16 8283133103 7 1.000000-5 0.000000+0 2.700000+6 7.200000-2 5.000000+6 1.800000-2283133103 8 7.000000+6 1.012500-2 1.000000+7 1.012500-2 1.200000+7 4.500000-3283133103 9 1.600000+7 1.800000-2 2.000000+7 0.000000+0 283133103 10 0.000000+0 0.000000+0 0 8 16 8283133103 11 1.000000-5 0.000000+0 2.700000+6 4.66640-17 5.000000+6 1.240600-7283133103 12 7.000000+6 2.178000-6 1.000000+7 1.147600-5 1.200000+7 9.112500-6283133103 13 1.600000+7 1.928300-5 2.000000+7 0.000000+0 283133103 14 283133 099999 2831 0 0 0 0 0 0 0 2.90630E+4 6.23890E+1 0 0 34 102925 1451 1 0.0 0.0 0 0 0 62925 1451 2 1.00000E+0 2.00000E+7 0 0 10 20022925 1451 3 3.00000E+2 0.0 1 0 550 82925 1451 4 29-Cu- 63 LANL,ORNL,FEI EVAL-FEB98 A.KONING ET AL,K.Zolotarev 2925 1451 5 DIST-Feb2004 2925 1451 6 ----IRDF-2002 MATERIAL 2925 2925 1451 7 -----INCIDENT NEUTRON DATA 2925 1451 8 ------ENDF-6 FORMAT 2925 1451 9 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2925 1451 10 DOSIMETRY 2925 1451 11 ******************************************************************2925 1451 12 29-Cu- 63 LANL,ORNL EVAL-FEB98 A.KONING,M.CHADWICK,HETRICK 2925 1451 13 Ch98,Ch99 DIST-SEP 1 REV4- 20010926 2925 1451 14 ----ENDF/B-VI MATERIAL 2925 REVISION 3 2925 1451 15 **************************************************************** 2925 1451 16 ***************************************************************** 2925 1451 17 29-Cu- 63 FEI EVAL-Nov01 K.I.Zolotarev 2925 1451 18 DIST-Jan02 20020110 2925 1451 19 ----BROND-2 MATERIAL 2925 2925 1451 20 -----INCIDENT NEUTRON DATA 2925 1451 21 ------ENDF-6 FORMAT 2925 1451 22 ------Russian Reactor Dosimetry File RRDF-2002 2925 1451 23 ***************************************************************** 2925 1451 24 Author of evaluation: K.I.Zolotarev 2925 1451 25 ----- MF=3 MT=107 ----- 2925 1451 26 ***************************************************************** 2925 1451 27 ******** Start of (N,2N), (N.G) bibliographical component ******* 2925 1451 28 ***************************************************************** 2925 1451 29 2925 1451 30 ENDF/B-VI MOD 5 Revision, May 2000, S.C. Frankle, R.C. Reedy, 2925 1451 31 P.G. Young (LANL) 2925 1451 32 2925 1451 33 The secondary gamma-ray spectrum for radiative capture (MF 12, 2925 1451 34 MT 102) has been updated for new experimental data at incident 2925 1451 35 neutron energies up to 1 keV. The Q-value for radiative capture 2925 1451 36 was also updated in File 3. 2925 1451 37 Details of these changes are described in Frankel et al. [Fr01]. 2925 1451 38 2925 1451 39 **************************************************************** 2925 1451 40 2925 1451 41 ENDF/B-VI MOD 4 Evaluation, February 1998, A.J. Koning (ECN), 2925 1451 42 M.B. Chadwick, P.G. Young (LANL) 2925 1451 43 2925 1451 44 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code 2925 1451 45 in cooperation with ECN Petten. 2925 1451 46 2925 1451 47 This evaluation provides a complete representation of the 2925 1451 48 nuclear data needed for transport, damage, heating, 2925 1451 49 radioactivity, and shielding applications over the incident 2925 1451 50 neutron energy range from 1.0E-11 to 150 MeV. The discussion 2925 1451 51 here is divided into the region below and above 20 MeV. 2925 1451 52 2925 1451 53 INCIDENT NEUTRON ENERGIES < 20 MeV 2925 1451 54 2925 1451 55 Below 20 MeV the evaluation is based completely on the ENDF/B- 2925 1451 56 VI (Mod 3) evaluation by D. Hetrick, C.Y. Fu, and D. Larson. 2925 1451 57 2925 1451 58 INCIDENT NEUTRON ENERGIES > 20 MeV 2925 1451 59 2925 1451 60 The ENDF/B-VI Release 2 evaluation extends to 20 MeV and 2925 1451 61 includes cross sections and energy-angle data for all 2925 1451 62 significant reactions. The present evaluation utilizes a more 2925 1451 63 compact composite reaction spectrum representation above 20 MeV 2925 1451 64 in order to reduce the length of the file. No essential data for 2925 1451 65 applications is lost with this representation. 2925 1451 66 The evaluation above 20 MeV utilizes MF=6, MT=5 to represent 2925 1451 67 all reaction data. Production cross sections and emission 2925 1451 68 spectra are given for neutrons, protons, deuterons, tritons, 2925 1451 69 alpha particles, gamma rays, and all residual nuclides produced 2925 1451 70 (A>5) in the reaction chains. To summarize, the ENDF sections 2925 1451 71 with non-zero data above En = 20 MeV are: 2925 1451 72 2925 1451 73 MF=3 MT= 1 Total Cross Section 2925 1451 74 MT= 2 Elastic Scattering Cross Section 2925 1451 75 MT= 3 Nonelastic Cross Section 2925 1451 76 MT= 5 Sum of Binary (n,n') and (n,x) Reactions 2925 1451 77 2925 1451 78 MF=4 MT= 2 Elastic Angular Distributions 2925 1451 79 2925 1451 80 MF=6 MT= 5 Production Cross Sections and Energy-Angle 2925 1451 81 Distributions for Emission Neutrons, Protons, 2925 1451 82 Deuterons, Tritons, and Alphas; and Angle- 2925 1451 83 Integrated Spectra for Gamma Rays and Residual 2925 1451 84 Nuclei That Are Stable Against Particle Emission 2925 1451 85 2925 1451 86 The evaluation is based on nuclear model calculations that 2925 1451 87 have been benchmarked to experimental data, especially for n + 2925 1451 88 Cu65 and p + Cu65 reactions [Ch98]. We use the GNASH code system 2925 1451 89 [Yo92], which utilizes Hauser-Feshbach statistical, preequilib- 2925 1451 90 rium and direct-reaction theories. Spherical optical model 2925 1451 91 calculations are used to obtain particle transmission 2925 1451 92 coefficients for the Hauser-Feshbach calculations, as well as 2925 1451 93 for the elastic neutron angular distributions. 2925 1451 94 Cross sections and spectra for producing individual residual 2925 1451 95 nuclei are included for reactions. The energy-angle-correlations 2925 1451 96 for all outgoing particles are based on Kalbach systematics 2925 1451 97 [Ka88]. 2925 1451 98 A model was developed to calculate the energy distributions of 2925 1451 99 all recoil nuclei in the GNASH calculations [Ch96a]. The recoil 2925 1451 100 energy distributions are represented in the laboratory system in 2925 1451 101 MT=5, MF=6, and are given as isotropic in the lab system. All 2925 1451 102 other data in MT=5,MF=6 are given in the center-of-mass system. 2925 1451 103 This method of representation utilizes the LCT=3 option approved 2925 1451 104 at the November, 1996, CSEWG meeting. 2925 1451 105 Preequilibrium corrections were performed in the course of the 2925 1451 106 GNASH calculations using the exciton model of Kalbach [Ka77, 2925 1451 107 Ka85], validated by comparison with calculations using Feshbach, 2925 1451 108 Kerman, Koonin (FKK) theory [Ch93]. Discrete level data from 2925 1451 109 nuclear data sheets were matched to continuum level densities 2925 1451 110 using the formulation of Ignatyuk et al. [Ig75] and pairing and 2925 1451 111 shell parameters from the Cook [Co67] analysis. Neutron and 2925 1451 112 charged- particle transmission coefficients were obtained from 2925 1451 113 the optical potentials, as discussed below. Gamma-ray 2925 1451 114 transmission coefficients were calculated using the Kopecky-Uhl 2925 1451 115 model [Ko90]. 2925 1451 116 2925 1451 117 SPECIFIC INFORMATION CONCERNING THE 63Cu EVALUATION 2925 1451 118 2925 1451 119 This evaluation is documented in some detail in Ref. [Ko98b]. 2925 1451 120 2925 1451 121 The neutron total cross section above 20 MeV was obtained by 2925 1451 122 evaluating experimental data, with a particular emphasis on the 2925 1451 123 Finlay [Fi93] elemental data. This resulted in an evaluated 2925 1451 124 elemental Cu total cross section; to obtain an isotopic 63Cu total2925 1451 125 cross section, it was assumed that 63Cu and 65Cu have total cross 2925 1451 126 sections in an A**2/3 ratio to one another. The total neutron 2925 1451 127 nonelastic cross section was obtained directly from an optical 2925 1451 128 model calculation (see below), after verifying that it was in good2925 1451 129 agreement with the experimental data [Ko98b]. 2925 1451 130 2925 1451 131 To obtain the neutron optical potential we used total cross 2925 1451 132 section data from 1.2 to 4.5 MeV [Gu86] and from 5.3 to 600 MeV 2925 1451 133 [Fi93], and elastic scattering angular distribution data from 1.6 2925 1451 134 to 96 MeV [Br50, Sa60, Ki74, El82, Gu86]. The optical potential 2925 1451 135 parameters were obtained using a combination of a grid search code2925 1451 136 and the interactive optical model viewer ECISVIEW [Ko97], both 2925 1451 137 built around the coupled channels code ECIS96 [Ra94]. The energy 2925 1451 138 dependence of the optical model parameters is as described in 2925 1451 139 [Ko98]. This optical potential was used for the calculation, with 2925 1451 140 ECIS96, of neutron transmission coefficients and DWBA cross 2925 1451 141 sections for the entire energy region above 20 MeV. 2925 1451 142 2925 1451 143 Due to the lack of proton elastic scattering data in numerical 2925 1451 144 form, we used a combination of global optical models for the 2925 1451 145 proton channel. The Becchetti-Greenlees potential [Be69]was 2925 1451 146 adopted below 47 MeV, and the non-relativistic version of the 2925 1451 147 Madland potential [Ma88] above 47 MeV. At this particular energy 2925 1451 148 point the two potentials join smoothly. 2925 1451 149 2925 1451 150 For deuterons, the Lohr-Haeberli global potential [Lo74] was used;2925 1451 151 for alpha particles the Moyen potential (MacFadden-Satchler 2925 1451 152 [Ma66]) was used; and for tritons the Becchetti-Greenlees 2925 1451 153 potential [Be71] was used. The He-3 channel was ignored, due to 2925 1451 154 its small importance. 2925 1451 155 2925 1451 156 Following Delaroche et al. [De82], we adopted the weak-coupling 2925 1451 157 model for direct collective inelastic scattering for Cu-63, using 2925 1451 158 Ni-64 as a basis. For the calculation of the cross sections, 2925 1451 159 ECIS96 was used in DWBA mode. We used the following direct 2925 1451 160 transitions for Cu-63 (ground state 3/2- ) : 2925 1451 161 2925 1451 162 Jpi Ex(MeV) Deformation lengths 2925 1451 163 0.5- 0.669 Delta(2)=0.319 2925 1451 164 2.5- 0.962 Delta(2)=0.552 2925 1451 165 3.5- 1.327 Delta(2)=0.638 2925 1451 166 1.5- 1.547 Delta(2)=0.451 2925 1451 167 1.5- 3.382 Delta(3)=0.313 2925 1451 168 2.5- 3.632 Delta(3)=0.384 2925 1451 169 3.5- 3.882 Delta(3)=0.444 2925 1451 170 4.5- 4.132 Delta(3)=0.496 2925 1451 171 2925 1451 172 No measurements exist for neutron-induced emission spectra above 2925 1451 173 20 MeV for 63Cu. However, for Cu-65 there exists 25.7 MeV (n,xn) 2925 1451 174 data by Marcinkowski et al [Ma83]. This has been used to benchmark2925 1451 175 the Cu-65 data. Without adjusting any of the level density or pre-2925 1451 176 equilibrium parameters the GNASH calculation was in good agreement2925 1451 177 with these data. Hence we also adopted these parameters for the 2925 1451 178 whole energy region for Cu-63. 2925 1451 179 2925 1451 180 **************************************************************** 2925 1451 181 2925 1451 182 REFERENCES 2925 1451 183 2925 1451 184 [Ab93] W. Abfalterer, R.W. Finlay, S.M. Grimes, and V. Mishra, 2925 1451 185 Phys.Rev. C47, 1033 (1993) 2925 1451 186 [Al83] R. Alarcon and J. Rapaport, Nucl.Phys. A458, 502 (1986) 2925 1451 187 [Ar80] E.D. Arthur and P.G. Young, 'Evaluation of Neutron Cross 2925 1451 188 Sections to 40 MeV for 54,56Fe," Proc. Sym. on Neutron Cross 2925 1451 189 Sections from 10 to 50 MeV, 12-14 May 1980, Brookhaven National2925 1451 190 Laboratory [Eds. M. R. Bhat and S. Pearlstein, BNL-NCS- 51245, 2925 1451 191 1980] p. 731. 2925 1451 192 [Be69] F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182, 2925 1451 193 1190 (1969) 2925 1451 194 [Be71] F.D. Becchetti, Jr., and G.W. Greenlees in "Polarization 2925 1451 195 Phenomena in Nuclear Reactions," (Ed: H.H. Barschall and W. 2925 1451 196 Haeberli, The University of Wisconsin Press, 1971) p.682. 2925 1451 197 [Be92] O. Bersillon, "SCAT2 - A Spherical Optical Model Code," 2925 1451 198 in Proc. ICTP Workshop on Computation and Analysis of Nuclear 2925 1451 199 data Relevant to Nuclear Energy and Safety, February-March, 2925 1451 200 1999 Trieste, Italy, to be published in World Scientific Press,2925 1451 201 and Progress Report of the Nuclear Physics Division, Bruyeres- 2925 1451 202 le-Chatel 1977, CEA-N-2037 (1978) p.111 2925 1451 203 [Br50] S. Bratenahl, S. Fernbach, R.H. Hildebrand et al., 2925 1451 204 Phys.Rev. 77, 597 (1950) 2925 1451 205 [Ch93] M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 2925 1451 206 [Ch96] M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J. 2925 1451 207 Koning, "High-Energy Nuclear Data Libraries for Accelerator- 2925 1451 208 Driven Technologies: Calculational Method for Heavy Recoils," 2925 1451 209 Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation 2925 1451 210 Technology and Applications, Kalmar, Sweden, 3-7 June 1996 2925 1451 211 [Ch98] M. B. Chadwick and P. G. Young, "GNASH Calculations of 2925 1451 212 n,p + Cu isotopes and Benchmarking of Results" in APT PROGRESS 2925 1451 213 REPORT: 1 February - 1 March 1998, internal Los Alamos National2925 1451 214 Laboratory memo, 6 Mar.1998 from R.E. MacFarlane to L. Waters. 2925 1451 215 [Ch99] M.B. Chadwick, P G. Young, G. M. Hale, et al., Los Alamos 2925 1451 216 National Laboratory report, LA-UR-99-1222 (1999) 2925 1451 217 [Co67] J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J. 2925 1451 218 Phys. 20, 477 (1967) 2925 1451 219 [De82] J.P. Delaroche, S.M. El-Kadi, P.P. Guss, C.E. Floyd and 2925 1451 220 R.L. Walter, Nucl. Phys. A390, 541 (1982). 2925 1451 221 [El82] S.M. El-Kadi, C.E. Nelson, F.O. Purser et al., Nucl.Phys. 2925 1451 222 A390, 509 (1982) 2925 1451 223 [Fi93] R. W. Finlay, W. P. Abfalterer, G. Fink et al., Phys.Rev. 2925 1451 224 C 47, 237 (1993) 2925 1451 225 [Fr01] S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos 2925 1451 226 National Laboratory Report, LA-13812 (2001). 2925 1451 227 [Gu86] P. Guenther, D.L. Smith, A.B. Smith, J.F. Whalen, Nucl. 2925 1451 228 Phys. A448, 280 (1986) 2925 1451 229 [Ig75] A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J. 2925 1451 230 Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485 2925 1451 231 (1975) 2925 1451 232 [Ka77] C. Kalbach, Z.Phys.A 283, 401 (1977) 2925 1451 233 [Ka85] C. Kalbach, Los Alamos National Laboratory report 2925 1451 234 LA-10248-MS (1985) 2925 1451 235 [Ka88] C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also 2925 1451 236 C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981) 2925 1451 237 [Ki74] W.E. Kinney, F.G. Perey, Oak Ridge report ORNL-4908 (1974)2925 1451 238 [Ko90] J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990) 2925 1451 239 [Ko97] A.J. Koning, J.J. van Wijk and J.-P. Delaroche, "ECISVIEW:2925 1451 240 A Graphical Interface for ECIS95", Proceedings of the NEA 2925 1451 241 Specialists' Meeting on the Nucleon Nucleus Optical Model up to2925 1451 242 200 MeV, Bruyeres-le-Chatel, November 13-15 1996. Available at 2925 1451 243 http://db.nea.fr/html/science/om200/. 2925 1451 244 [Ko98] A.J. Koning, J.-P. Delaroche and O. Bersillon, "Nuclear 2925 1451 245 Data for Accelerator-Driven Systems: Nuclear models, Experiment2925 1451 246 and Data Libraries", to appear in Mucl. Instr. Meth. A (1998). 2925 1451 247 [Ko98b] A.J. Koning, M.B. Chadwick, and P.G. Young, "ENDF/B-VI 2925 1451 248 neutron and proton datafiles up to 150 MeV for 63Cu and 65Cu", 2925 1451 249 Los Alamos National Laboratory report LAUR- (1998); ECN lab and2925 1451 250 JEFF report (1998). 2925 1451 251 [Lo74] J.M. Lohr and W. Haeberli, Nucl.Phys. A232, 381 (1974) 2925 1451 252 [Ma66] Macfadden and Satchler, Nuc.Phys. 84, 177 (1966) 2925 1451 253 [Ma83] A. Marcinkowski, R.W. Finlay, G. Randers-Pehrson et al., 2925 1451 254 Nucl.Phys. A402, 220 (1983) 2925 1451 255 [Ma88] D.G. Madland, "Recent Results in the Development of a 2925 1451 256 Global Medium-Energy Nucleon-Nucleus Optical-Model Potential, 2925 1451 257 "Proc. OECD/NEANDC Specialist's Mtg. on Preequilibrium Nuclear 2925 1451 258 Reactions, Semmering, Austria, 10-12 Feb. 1988, NEANDC-245 'U' 2925 1451 259 (1988). 2925 1451 260 [Pe63] C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963) 2925 1451 261 [Ra94] J. Raynal, Notes on ECIS94, CEA Saclay Report CEA-N-2772 2925 1451 262 (1994) 2925 1451 263 [Sa60] G.L. Salmon, Nucl.Phys. 21, 15 (1960) 2925 1451 264 [Yo92] P.G. Young, E.D. Arthur, and M.B. Chadwick, report 2925 1451 265 LA-12343-MS (1992) 2925 1451 266 2925 1451 267 **************************************************************** 2925 1451 268 2925 1451 269 ENDF/B-VI MOD 3 Revision, July 1991 (ORNL) 2925 1451 270 2925 1451 271 MOD 3 changes 2925 1451 272 1) Corrections to MF=6, MT=65 at 17.0 MeV to prevent negative 2925 1451 273 values in the angular distribution. 2925 1451 274 2) Corrections to MF=33, MT=102 2925 1451 275 2925 1451 276 **************************************************************** 2925 1451 277 2925 1451 278 * Note there was no MOD 2 released. 2925 1451 279 2925 1451 280 **************************************************************** 2925 1451 281 2925 1451 282 ENDF/B-VI MOD 1 Evaluation, October 1989, D. Hetrick, F.Y. Fu, 2925 1451 283 D. Larson (ORNL) 2925 1451 284 2925 1451 285 This work employed several nuclear model codes including the 2925 1451 286 optical-model code GENOA [1], the Distorted Wave Born 2925 1451 287 Approximation (DWBA) program DWUCK [2], and the Hauser-Feshbach 2925 1451 288 code TNG [3,4]. The TNG code provides energy and angular 2925 1451 289 distributions of particles emitted in the compound and pre- 2925 1451 290 compound reactions, ensures consistency among all reactions, and 2925 1451 291 maintains energy balance. Details pertinent to the contents of 2925 1451 292 this evaluation and extensive comparisons of calculations with 2925 1451 293 experimental data can be found in reference [5]. 2925 1451 294 2925 1451 295 ----- DESCRIPTION OF FILES 2925 1451 296 (MF-MT) 2925 1451 297 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2925 1451 298 2-151 RESONANCE PARAMETERS WERE TAKEN FROM MUGHABGHAB[6]. POINT2925 1451 299 WISE RECONSTRUCTION COMPARED WITH DATA [7] SHOWED POORER 2925 1451 300 FIT ABOVE 100 KEV, SO THE RESONANCE REGION WAS CUT OFF AT 2925 1451 301 99.5 KEV. REICH-MOORE PARAMETERS ARE GIVEN. AGREEMENT 2925 1451 302 WITH DATA COULD BE IMPROVED WITH ADDITION OF A BACKGROUND 2925 1451 303 FILE IN 3/1, BUT THIS IN GENERAL GIVES TOO LARGE AN 2925 1451 304 AVERAGE CROSS SECTION, WHEN BINNED IN 10 KEV BINS AND 2925 1451 305 COMPARED WITH THE BINNED DATA. THIS IS PROBABLY DUE TO 2925 1451 306 TOO LARGE AN ESTIMATE OF NEUTRON WIDTHS FOR RESONANCES 2925 1451 307 SEEN ONLY IN CAPTURE AND NOT IN TRANSMISSION. 2925 1451 308 NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2925 1451 309 OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE 2925 1451 310 PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON 2925 1451 311 A FINER ENERGY GRID THAN GIVEN IN 4/2. 2925 1451 312 3-1 THE TOTAL CROSS SECTION IS GIVEN BY RESONANCE PARAMETERS 2925 1451 313 FROM 1.E-5 EV TO 99.5 KEV. FROM 1.E-5 TO 1 EV, -0.9B IS 2925 1451 314 GIVEN TO REDUCE THE ELASTIC AND GIVE THE CORRECT TOTAL 2925 1451 315 THERMAL CROSS SECTION (9.6 B). FROM 1 EV TO 170 EV THIS 2925 1451 316 GOES LINEARLY TO ZERO. A SMALL CONTRIBUTION FROM 3/102 IS2925 1451 317 REQUIRED FROM 60 TO 99.5 KEV, WHICH WHEN ADDED TO 2/151 2925 1451 318 REPRODUCES THE CAPTURE DATA. FROM 99.5 KEV TO 1.12 MEV 2925 1451 319 CU63 DATA FROM [7] IS USED, AFTER APPROPRIATE AVERAGING. 2925 1451 320 ABOVE THIS, NO ISOTOPIC DATA IS AVAILABLE. FROM 1.12 TO 2925 1451 321 4.0 MEV,NAT CU DATA OF PEREY [8] USED IN V5 IS RETAINED. 2925 1451 322 FROM 4.0 TO 20 MEV, NAT CU DATA OF LARSON ET.AL [9] IS 2925 1451 323 AVERAGED AND USED. COMPARISONS FROM 1.2 TO 4.5 2925 1451 324 MEV WITH AVERAGED ARGONNE DATA FOR NAT CU [10] SHOW 2925 1451 325 1% AGREEMENT. 2925 1451 326 3-2 ELASTIC SCATTERING CROSS SECTIONS WERE OBTAINED BY 2925 1451 327 SUBTRACTING THE NONELASTIC (3-3) FROM THE TOTAL. THE 2925 1451 328 THERMAL VALUE OF 5.1 B IS REPRODUCED. 2925 1451 329 3-102 (N,G) DATA TAKEN FROM RESONANCE PARAMETERS FROM 1.E-5 EVTO2925 1451 330 99.5 KEV. A SMALL BACKGROUND IS GIVEN HERE WHICH WHEN 2925 1451 331 ADDED TO THE RESONACE CONTRIBUTION REPRODUCES EXPERIMENTAL2925 1451 332 DATA,INCLUDING THE THERMAL VALUE OF 4.50B. V5 DATA FROM 2925 1451 333 99.5 KEV TO 20.0 MEV WERE REPLACED BY POINTS ON A CURVE 2925 1451 334 DRAWN THROUGH DATA FROM THE CSISRS LIBRARY [5,13]; RESULTS2925 1451 335 COMPARABLE TO EYE GUIDE IN REF [14]. 2925 1451 336 ---------------------------------------------------------------- 2925 1451 337 UNCERTAINTY FILES 2925 1451 338 ALL NON-DERIVED FILES CONTAIN AN LB=8 COMPONENT, AS 2925 1451 339 REQUIRED BY ENDF/B-VI FORMATS 2925 1451 340 2925 1451 341 33-1 TOTAL UNCERTAINTIES GIVEN AS DERIVED FROM 1E-5 TO 200 EV 2925 1451 342 EXPLICIT FROM 200 EV TO 20 MEV, USING LB=0,1 AND 8. 2925 1451 343 33-2 EXPLICIT FROM 1E-5 T0 200 EV, DERIVED FROM 200EV TO 20 MEV2925 1451 344 33-102 CAPTURE UNCERTAINTIES ESTIMATED FROM THERMAL VALUE AT LOW 2925 1451 345 ENERGIES, BINNED DATA IN THE RESONANCE REGION, AND CSISRS 2925 1451 346 DATA [5,13,14] FROM 99.5 KEV TO 20 MEV. 2925 1451 347 **************************************************************** 2925 1451 348 2925 1451 349 REFERENCES: 2925 1451 350 2925 1451 351 [1] F.G. Perey, computer code GENOA, ORNL, unpublished (1967) 2925 1451 352 [2] P.D. Kunz, "Distorted Wave Code DWUCK72," Univ. of 2925 1451 353 Colorado, unpublished (1972) 2925 1451 354 [3] C.Y. Fu, report ORNL/TM-7042 (1980); also, C.Y Fu, 2925 1451 355 Symp. on Neutron Cross Sections from 10 to 50 MeV, Upton, NY,2925 1451 356 May 1980, Brookhaven National Lab. report BNL-NCS-51245 2925 1451 357 (1980) p.675 2925 1451 358 [4] K. Shibata and C.Y. Fu, report ORNL/TM-10093 (1986) 2925 1451 359 [5] D.M. Hetrick, C.Y. Fu, and D.C. Larson, Oak Ridge report 2925 1451 360 ORNL/TM-9083 [ENDF-337] (1984) 2925 1451 361 [6] S.F. Mughabghab, M. Divadeenam, and N.E. Holden, "Neutron 2925 1451 362 Cross Sections, Vol. 1, Neutron Resonance Parameters and 2925 1451 363 Thermal Cross Sections, Part A, Z=1-60," (Academic Press, 2925 1451 364 1981) 2925 1451 365 [7] M.S. Pandey, J.B. Garg and J.A. Harvey, Phys.Rev. C 15, 600 2925 1451 366 (1977), and private communication. 2925 1451 367 [8] F.G. Perey, private communication (1977) 2925 1451 368 [9] D.C. Larson, Symp. on Neutron Cross Sections from 10 to 50 2925 1451 369 MeV, Upton, NY, May 1980, Brookhaven National Lab. report 2925 1451 370 BNL-NCS-51245 (1980) p.277 2925 1451 371 [10] P. Guenther, D.L. Smith, A.B. Smith and J.F. Whalen, Nucl. 2925 1451 372 Phys. A, 448, 280 (1986) [CSISRS data set 12869/002], and 2925 1451 373 W.P. Poenitz and J.F. Whalen, Argonne report ANL/NDM-80 2925 1451 374 (1983) [CSISRS data set 12853] 2925 1451 375 [11] D.M. Hetrick and C.Y. Fu, Oak Ridge report ORNL/TM-7341 2925 1451 376 [ENDF-303] (1980) 2925 1451 377 [12] C.Y. Fu and D.M. Hetrick, Proc. Fourth ASTM-Euratom Symp. 2925 1451 378 on Reactor Dosimetry, Gaithersburg, Maryland, March 22-26, 2925 1451 379 1982 (U.S. National Bureau of Standards) p.877 2925 1451 380 [13] CSISRS Library, National Nuclear Data Center, Brookhaven 2925 1451 381 National Laboratory, Upton, N.Y. 11973. 2925 1451 382 [14] V. McLane, C.L. Dunford and P.F. Rose, "Neutron Cross 2925 1451 383 Sections, Vol. 2, Neutron Cross Section Curves" (Academic 2925 1451 384 Press, 1988) 2925 1451 385 [15] S.M. Qaim, Radiochimica Acta, 25, 13 (1978) 2925 1451 386 [16] M.G. Delfini, J. Kopecky, R.E. Chrien et al., Nucl.Phys. 2925 1451 387 A404, 250 (1983) 2925 1451 388 2925 1451 389 ***************************************************************** 2925 1451 390 ******** End of (N,2N), (N.G) bibliographical component ******* 2925 1451 391 ***************************************************************** 2925 1451 392 ***************************************************************** 2925 1451 393 ******** Start of (N,A) bibliographical component ******* 2925 1451 394 ***************************************************************** 2925 1451 395 ------Russian Reactor Dosimetry File RRDF-2002 2925 1451 396 ***************************************************************** 2925 1451 397 Author of evaluation: K.I.Zolotarev 2925 1451 398 ***************************************************************** 2925 1451 399 2925 1451 400 ----- MF=3 MT=107 ----- 2925 1451 401 Evaluation of Cu-63(n,a)Co-60m+g- excitation function was car-2925 1451 402 ried out by means of statistical analysis of cross sections from 2925 1451 403 data base prepared in the energy range 2 - 20 Mev. In the energy 2925 1451 404 range 3.56 - 19.55 MeV input data base was formed with using of 2925 1451 405 experimental data from ref. [1-19]. Cross section data in the 2925 1451 406 interval 2.0 - 3.5 MeV were taken from theoretical model calcula- 2925 1451 407 tion. Experimental data included in the input data base were 2925 1451 408 renormalized using new cross sections standards for monitor reac- 2925 1451 409 tions and new standards for decay data. 2925 1451 410 The special correction was applied to the experimental data 2925 1451 411 [2,3,12,13,15,18]. Cross section data of A.Paulsen and H.Liskien 2925 1451 412 [2-3] measured in the energy region 12.09 - 19.55 MeV with using 2925 1451 413 T(d,n)He4 neutron source were multiplied to the factor 1.20805 . 2925 1451 414 Experimental data of Lu Hanlin et al. [12,18] , Wang Yongchag et 2925 1451 415 al. [13] and Konno et al. [15] were multiplied to the factors 2925 1451 416 0.88110, 0.86000, 0.84377 and 1.060, respectively. The correction 2925 1451 417 factors were derived from preliminary evaluated cross sections 2925 1451 418 integrals in the energy intervals 8.4 - 11.4 and 13 - 15 MeV . 2925 1451 419 Data of A.Paulsen and H.Liskien [2] in the energy range 2925 1451 420 5.76 - 11.48 MeV measured with using D(d,n)He3 , Be9(a,n)C12 , 2925 1451 421 C14(d,n)N15 , N15(d,n)O16 neutron sources were rejected due to 2925 1451 422 their inconsistency with precision measurements of G.Winkler et 2925 1451 423 al. [8] and integral experimental data for U-235 fission neutron 2925 1451 424 spectrum [20-23] and Cf-252 spontaneous fission neutron spectrum 2925 1451 425 [25-26]. Cross sections measured by J.Kantele and D.Gardner [27], 2925 1451 426 M.Bormann et al. [28], G.Maslov et al. [29] and K.Kayashima et al.2925 1451 427 [30] were also rejected due to the big discrepancy with the main 2925 1451 428 bulk experimental data. 2925 1451 429 The final procedure evaluation of (n,a) excitation function 2925 1451 430 was carried out by means of Pade-2 code [31]. 2925 1451 431 Evaluated excitation function for the reaction Cu63(n,a)Co60 2925 1451 432 was tested with using integral experimental data [20-24] for 2925 1451 433 U-235 thermal fission neutron spectrum and evaluated integral ex- 2925 1451 434 perimental data [32-33] for Cf-252 spontaneous fission neutron 2925 1451 435 spectrum. Calculated and measured average cross section values 2925 1451 436 for U-235 thermal fission neutron spectrum [34] and Cf-252 sponta-2925 1451 437 neous fission neutron spectrum [35] are given in the table 1. 2925 1451 438 Table 1 2925 1451 439 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2925 1451 440 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 2925 1451 441 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2925 1451 442 U-235 neutron fission ³ 0.53294 ³ 0.5269 +- 0.0316 [20] 2925 1451 443 ³ ³ 0.520 +- 0.040 [21] 2925 1451 444 ³ ³ 0.519 +- 0.035 [22] 2925 1451 445 ³ ³ 0.5273 +- 0.0156 [23] 2925 1451 446 ³ ³ 0.538 +- 0.015 [23] 2925 1451 447 ³ ³ 0.4935 +- 0.0242 [24] 2925 1451 448 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2925 1451 449 CF-252 spont. fission ³ 0.69248 ³ 0.709 +- 0.017 [25] 2925 1451 450 ³ ³ 0.675 +- 0.018 [26] 2925 1451 451 ³ ³ 0.6897 +- 0.0130 [32] 2925 1451 452 ³ ³ 0.6887 +- 0.0135 [33] 2925 1451 453 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 2925 1451 454 2925 1451 455 ----- MF=33 MT=107 ----- 2925 1451 456 Uncertainties in the evaluated excitation function for the 2925 1451 457 reaction Cu-63(n,a)Co-60m+g are given in the form of relative co- 2925 1451 458 variance matrix for the 18-neutron energy groups (LB=5). Covari- 2925 1451 459 ance matrix of uncertainties was calculated simultaneously with 2925 1451 460 recommended cross section data by means of PADE-2 code. 2925 1451 461 Eigenvalues of the 6-th digits relative covariance matrix 2925 1451 462 given in the 33-file are the following: 2925 1451 463 2925 1451 464 9.39333E-10 1.02943E-09 1.18681E-09 1.49652E-09 2925 1451 465 2.03282E-09 2.85265E-09 3.69906E-09 6.46604E-09 2925 1451 466 7.96491E-09 7.69999E-08 4.12497E-06 6.99817E-04 2925 1451 467 1.07280E-03 1.33316E-03 1.89667E-03 4.15078E-03 2925 1451 468 6.87556E-03 8.81528E-03 2925 1451 469 2925 1451 470 References : 2925 1451 471 1. B.Czapp, H.Vonach Oesterr. Akad. Wiss, Math, Naturw. 2925 1451 472 Anzeiger, v.97, p.13, January 1960 2925 1451 473 2. A.Paulsen, H.Liskien Nukleonik, v.10, p.91, July 1967 2925 1451 474 3. A.Paulsen Zeitschrift f. Phys., v.205, p.226, August 1967 2925 1451 475 4. R.C.Barrall et al. Report AFWL-TR-68-134, March 1969 2925 1451 476 6. G.Winkler Nucl. Sci. Eng., v.67, n.2, p.260, August 1978 2925 1451 477 8. G.Winkler, D.L.Smith, J.W.Meadows Nucl. Sci. Eng., v.76, 2925 1451 478 p.30, October 1980 2925 1451 479 9. U.Garuska et al. Prog. Report INDC(POL)-11, p.15, July 1980 2925 1451 480 10. O.I.Artem'ev et al. Atomnaja Energija (Sov.), v.49, n.3, 2925 1451 481 p.195, September 1980 2925 1451 482 11. L.R.Greenwood Progress Report ASTM-STP-956, p.743, 1987 2925 1451 483 12. Lu Hanlin et al. China Journal of Nuclear Phys., v.12. n.4, 2925 1451 484 p. 373, 1990 2925 1451 485 13. Wang Yongchang et al. Chinese J. High Energy Phys. and Nucl. 2925 1451 486 Phys., v.14, p.919, October 1990 2925 1451 487 14. J.Csikai, C.M.Buczko, R.Pepelnik, H.M.Agrawal Annals of Nucl. 2925 1451 488 Energy, v.18, n.1, p.1, 1991 2925 1451 489 15. C.Konno et al. Report JAERI-1329, October 1993 2925 1451 490 16. J.W.Meadows et al. Annals of Nucl. Energy, v.23, p.877, 1996 2925 1451 491 17. A.A.Filatenkov et al. VANT, Ser.: Yadernye Konstanty, v.2, 2925 1451 492 p.8, Moscow, 1996 2925 1451 493 18. Lu Hanlin et al. Report INDC(CPR)-045, IAEA, October 1998 2925 1451 494 19. A.A.Filatenkov et al. Report RI-252, St.Petersburg, May 1999 2925 1451 495 20. R.Lloret Progress Report EANDC(E)-57, p.172, February 1965 2925 1451 496 21. A.Fabry, J.P.Deworm Progress Report EANDC(E)-66, p.125, 2925 1451 497 February 1966 2925 1451 498 22. K.Kobayashi et al. Nucl. Sci. Techn., v.13, p.531, Oct. 1976 2925 1451 499 23. L.P.Geraldo et al. Radiochimica Acta, v.57, pp.63-67, 1992 2925 1451 500 24. W.Mannhart Progress Report INDC(Ger)-045, pp.40-43, 1999 2925 1451 501 25. G.Winkler et al. Nucl. Sci. Eng., v.78, p.415, August 1981 2925 1451 502 26. W.Mannhart Proc. of Int. Conf. Nuclear Data for Science and 2925 1451 503 Technology, 6-10 September 1982, Antwerp, Holland, D.Reidel 2925 1451 504 Publishing Company, p.429 2925 1451 505 27. J.Kantele, D.Gardner Nucl. Phys., v.35, p.353, 1962 2925 1451 506 28. M.Bormann et al. Nucl. Phys., v.A186, p.65, May 1972 2925 1451 507 29. G.N.Maslov, F.Nasyrov, N.F.Pashkin Yadernye Konstanty, v.9, 2925 1451 508 p.50, Obninsk, 1972 2925 1451 509 30. K.Kayashima et al. Prog. Rep. NEANDC(J)-61U, p.94, Sep. 1979 2925 1451 510 31. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 2925 1451 511 32. W.Mannhart Handbook on Nuclear Activation Cross Sections, 2925 1451 512 IAEA Technical Report series No.273, p.413, 1987 2925 1451 513 33. W.Mannhart Validation of Differential Cross Sections with 2925 1451 514 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 2925 1451 515 September 2002 2925 1451 516 34. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 2925 1451 517 Library, MAT=9228, MF=5, MT=18, eval. April 1989 2925 1451 518 35. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 2925 1451 519 ***************************************************************** 2925 1451 520 ******** End of (N,A)bibliographical component ******* 2925 1451 521 ***************************************************************** 2925 1451 522 2925 1451 523 2925 1451 524 2925 1451 525 ************************ C O N T E N T S *********************** 2925 1451 526 ***************** Program LINEAR (VERSION 2002-1) ***************2925 1451 527 For All Data Greater than 1.0000E-10 barns in Absolute Value 2925 1451 528 Data Linearized to Within an Accuracy of .100000000 per-cent 2925 1451 529 ***************** Program RECENT (VERSION 2002-1) ***************2925 1451 530 for All Data Greater than 1.0000E-10 barns in Absolute Value 2925 1451 531 Data Linearized to within an Accuracy of .100000000 per-cent 2925 1451 532 ***************** Program SIGMA1 (VERSION 2002-1) ***************2925 1451 533 Data Doppler Broadened to 300.000000 Kelvin 2925 1451 534 for All Data Greater than 1.0000E-10 barns in Absolute Value 2925 1451 535 Data Linearized to Within an Accuracy pf .100000000 per-cent 2925 1451 536 ***************** Program FIXUP (Version 2002-1) ****************2925 1451 537 Corrected ZA/AWR in All Sections-----------------------------Yes 2925 1451 538 Corrected Thresholds-----------------------------------------Yes 2925 1451 539 Extended Cross Sections to 20 MeV----------------------------No 2925 1451 540 Allow Cross Section Deletion---------------------------------No 2925 1451 541 Allow Cross Section Reconstruction---------------------------No 2925 1451 542 Make All Cross Sections Non-Negative-------------------------Yes 2925 1451 543 Delete Energies Not in Ascending Order-----------------------Yes 2925 1451 544 Deleted Duplicate Points-------------------------------------Yes 2925 1451 545 Check for Ascending MAT/MF/MT Order--------------------------Yes 2925 1451 546 Check for Legal MF/MT Numbers--------------------------------Yes 2925 1451 547 Allow Creation of Missing Sections---------------------------No 2925 1451 548 Allow Insertion of Energy Points-----------------------------No 2925 1451 549 Create Uniform Energy Grid-----------------------------------No 2925 1451 550 Delete Section if Cross Section =0 at All Energies-----------Yes 2925 1451 551 ***************** Program GROUPIE (VERSION 2002-1) **************2925 1451 552 Unshielded Group Averages Using 640 Groups 2925 1451 553 Weighting Spectrum: Flat (Constant) Spectrum 2925 1451 554 1 451 562 52925 1451 555 2 151 4 12925 1451 556 3 16 34 12925 1451 557 3 102 217 52925 1451 558 3 107 63 12925 1451 559 33 16 58 12925 1451 560 33 102 23 32925 1451 561 33 107 38 12925 1451 562 2925 1 099999 2925 0 0 0 2.90630E+4 6.23890E+1 0 0 1 02925 2151 1 2.90630E+4 1.00000E+0 0 0 1 02925 2151 2 1.00000E-5 9.95000E+4 0 0 0 02925 2151 3 1.50000E+0 6.70000E-1 0 0 0 02925 2151 4 2925 2 099999 2925 0 0 0 2.90630E+4 6.23890E+1 0 0 0 02925 3 16 1 -1.08540E+7-1.08540E+7 0 0 1 912925 3 16 2 91 1 2925 3 16 3 11000000.0 .000429103 11100000.0 .002018629 11200000.0 .0036728762925 3 16 4 11300000.0 .006750000 11400000.0 .011250000 11500000.0 .0182500002925 3 16 5 11600000.0 .027750000 11700000.0 .039500000 11800000.0 .0535000002925 3 16 6 11900000.0 .067948571 12000000.0 .082845714 12100000.0 .0977428572925 3 16 7 12200000.0 .113333107 12300000.0 .133082000 12400000.0 .1535240002925 3 16 8 12500000.0 .173966000 12600000.0 .194408000 12700000.0 .2150142502925 3 16 9 12800000.0 .236606000 12900000.0 .258362000 13000000.0 .2801180002925 3 16 10 13100000.0 .301874000 13200000.0 .322831375 13300000.0 .3389970002925 3 16 11 13400000.0 .354364000 13500000.0 .369731000 13600000.0 .3850980002925 3 16 12 13700000.0 .400465000 13800000.0 .415832000 13900000.0 .4311990002925 3 16 13 14000000.0 .446566000 14100000.0 .461933000 14200000.0 .4772576252925 3 16 14 14300000.0 .492328000 14400000.0 .507356000 14500000.0 .5223840002925 3 16 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1.50000E+8 0 0 10 20022931 1451 3 3.00000E+2 0.0 1 0 384 42931 1451 4 29-Cu- 65 LANL,ORNL EVAL-FEB98 A.KONING,M.CHADWICK,HETRICK 2931 1451 5 DIST-Feb2002 2931 1451 6 ----IRDF-2002 MATERIAL 2931 2931 1451 7 -----INCIDENT NEUTRON DATA 2931 1451 8 ------ENDF-6 FORMAT 2931 1451 9 ******************************************************************2931 1451 10 29-Cu- 65 LANL,ORNL EVAL-FEB98 A.KONING,M.CHADWICK,HETRICK 2931 1451 11 Ch98,Ch99 DIST-SEP 1 REV4- 20010926 2931 1451 12 ----ENDF/B-VI MATERIAL 2931 REVISION 4 2931 1451 13 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************2931 1451 14 DOSIMETRY 2931 1451 15 ******************************************************************2931 1451 16 2931 1451 17 **************************************************************** 2931 1451 18 2931 1451 19 ENDF/B-VI MOD 5 Revision, May 2000, S.C. Frankle, R.C. Reedy, 2931 1451 20 P.G. Young (LANL) 2931 1451 21 2931 1451 22 The secondary gamma-ray spectrum for radiative capture (MF 12, 2931 1451 23 MT 102) has been updated for new experimental data at incident 2931 1451 24 neutron energies up to 1 keV. 2931 1451 25 The MF=12, MT=102 yields above 1 keV were adjusted slightly to 2931 1451 26 force energy conservation. 2931 1451 27 The Q-value for radiative capture was also updated in File 3. 2931 1451 28 Details of these changes are described in Frankel et al. [Fr01]. 2931 1451 29 2931 1451 30 **************************************************************** 2931 1451 31 2931 1451 32 ENDF/B-VI MOD 4 Evaluation, February 1998, A.J. Koning (ECN), 2931 1451 33 M.B. Chadwick, P.G. Young (LANL) 2931 1451 34 2931 1451 35 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code 2931 1451 36 in cooperation with ECN Petten. 2931 1451 37 2931 1451 38 This evaluation provides a complete representation of the 2931 1451 39 nuclear data needed for transport, damage, heating, 2931 1451 40 radioactivity, and shielding applications over the incident 2931 1451 41 neutron energy range from 1.0E-11 to 150 MeV. The discussion 2931 1451 42 here is divided into the region below and above 20 MeV. 2931 1451 43 2931 1451 44 INCIDENT NEUTRON ENERGIES < 20 MeV 2931 1451 45 2931 1451 46 Below 20 MeV the evaluation is based completely on the ENDF/B- 2931 1451 47 VI (Release 2) evaluation by D. Hetrick, C.Y. Fu, and D. Larson. 2931 1451 48 2931 1451 49 INCIDENT NEUTRON ENERGIES > 20 MeV 2931 1451 50 2931 1451 51 The ENDF/B-VI Release 2 evaluation extends to 20 MeV and 2931 1451 52 includes cross sections and energy-angle data for all 2931 1451 53 significant reactions. The present evaluation utilizes a more 2931 1451 54 compact composite reaction spectrum representation above 20 MeV 2931 1451 55 in order to reduce the length of the file. No essential data for 2931 1451 56 applications is lost with this representation. 2931 1451 57 The evaluation above 20 MeV utilizes MF=6, MT=5 to represent 2931 1451 58 all reaction data. Production cross sections and emission 2931 1451 59 spectra are given for neutrons, protons, deuterons, tritons, 2931 1451 60 alpha particles, gamma rays, and all residual nuclides produced 2931 1451 61 (A>5) in the reaction chains. To summarize, the ENDF sections 2931 1451 62 with non-zero data above En = 20 MeV are: 2931 1451 63 2931 1451 64 MF=3 MT= 1 Total Cross Section 2931 1451 65 MT= 2 Elastic Scattering Cross Section 2931 1451 66 MT= 3 Nonelastic Cross Section 2931 1451 67 MT= 5 Sum of Binary (n,n') and (n,x) Reactions 2931 1451 68 2931 1451 69 MF=4 MT= 2 Elastic Angular Distributions 2931 1451 70 2931 1451 71 MF=6 MT= 5 Production Cross Sections and Energy-Angle 2931 1451 72 Distributions for Emission Neutrons, Protons, 2931 1451 73 Deuterons, Tritons, and Alphas; and Angle- 2931 1451 74 Integrated Spectra for Gamma Rays and Residual 2931 1451 75 2931 1451 76 Nuclei That Are Stable Against Particle Emission 2931 1451 77 2931 1451 78 The evaluation is based on nuclear model calculations that 2931 1451 79 have been benchmarked to experimental data, especially for n + 2931 1451 80 Cu65 and p + Cu65 reactions (Ch98). We use the GNASH code system 2931 1451 81 (Yo92), which utilizes Hauser-Feshbach statistical, preequilib- 2931 1451 82 rium and direct-reaction theories. Spherical optical model 2931 1451 83 calculations are used to obtain particle transmission 2931 1451 84 coefficients for the Hauser-Feshbach calculations, as well as 2931 1451 85 for the elastic neutron angular distributions. 2931 1451 86 Cross sections and spectra for producing individual residual 2931 1451 87 nuclei are included for reactions. The energy-angle-correlations 2931 1451 88 for all outgoing particles are based on Kalbach systematics 2931 1451 89 (Ka88). 2931 1451 90 A model was developed to calculate the energy distributions of 2931 1451 91 all recoil nuclei in the GNASH calculations (Ch96a). The recoil 2931 1451 92 energy distributions are represented in the laboratory system in 2931 1451 93 MT=5, MF=6, and are given as isotropic in the lab system. All 2931 1451 94 other data in MT=5,MF=6 are given in the center-of-mass system. 2931 1451 95 This method of representation utilizes the LCT=3 option approved 2931 1451 96 at the November, 1996, CSEWG meeting. 2931 1451 97 Preequilibrium corrections were performed in the course of the 2931 1451 98 GNASH calculations using the exciton model of Kalbach (Ka77, 2931 1451 99 Ka85), validated by comparison with calculations using Feshbach, 2931 1451 100 Kerman, Koonin (FKK) theory [Ch93]. Discrete level data from 2931 1451 101 nuclear data sheets were matched to continuum level densities 2931 1451 102 using the formulation of Ignatyuk et al. (Ig75) and pairing and 2931 1451 103 shell parameters from the Cook (Co67) analysis. Neutron and 2931 1451 104 charged- particle transmission coefficients were obtained from 2931 1451 105 the optical potentials, as discussed below. Gamma-ray 2931 1451 106 transmission coefficients were calculated using the Kopecky-Uhl 2931 1451 107 model (Ko90). 2931 1451 108 2931 1451 109 SPECIFIC INFORMATION CONCERNING THE 65Cu EVALUATION 2931 1451 110 2931 1451 111 This evaluation is documented in some detail in Ref. (Ko98b). 2931 1451 112 2931 1451 113 The neutron total cross section above 20 MeV was obtained by 2931 1451 114 evaluating experimental data, with a particular emphasis on the 2931 1451 115 Finlay (Fi93) elemental data. This resulted in an evaluated 2931 1451 116 elemental Cu total cross section; to obtain an isotopic 65Cu total2931 1451 117 cross section, it was assumed that 63Cu and 65Cu have total cross 2931 1451 118 sections in an A**2/3 ratio to one another. The total neutron 2931 1451 119 nonelastic cross section was obtained directly from an optical 2931 1451 120 model calculation (see below), after verifying that it was in good2931 1451 121 agreement with the experimental data (Ko98b). 2931 1451 122 2931 1451 123 To obtain the neutron optical potential we used total cross 2931 1451 124 section data from 1.2 to 4.5 MeV (Gu86) and from 5.3 to 600 MeV 2931 1451 125 (Fi93), and elastic scattering angular distribution data from 1.6 2931 1451 126 to 96 MeV (Br50, Sa60, Ki74, El82, Gu86). The optical potential 2931 1451 127 parameters were obtained using a combination of a grid search code2931 1451 128 and the interactive optical model viewer ECISVIEW [Ko97], both 2931 1451 129 built around the coupled channels code ECIS96 [Ra94]. The energy 2931 1451 130 dependence of the optical model parameters is as described in 2931 1451 131 [Ko98]. This optical potential was used for the calculation, with 2931 1451 132 ECIS96, of neutron transmission coefficients and DWBA cross 2931 1451 133 sections for the entire energy region above 20 MeV. 2931 1451 134 2931 1451 135 Due to the lack of proton elastic scattering data in numerical 2931 1451 136 form, we used a combination of global optical models for the 2931 1451 137 proton channel. The Becchetti-Greenlees potential [Be69]was 2931 1451 138 adopted below 47 MeV, and the non-relativistic version of the 2931 1451 139 Madland potential [Ma88] above 47 MeV. At this particular energy 2931 1451 140 point the two potentials join smoothly. 2931 1451 141 2931 1451 142 For deuterons, the Lohr-Haeberli global potential [Lo74] was used;2931 1451 143 for alpha particles the Moyen potential (MacFadden-Satchler 2931 1451 144 [Ma66]) was used; and for tritons the Becchetti-Greenlees 2931 1451 145 potential [Be71] was used. The He-3 channel was ignored, due to 2931 1451 146 its small importance. 2931 1451 147 2931 1451 148 Following Delaroche et al. [De82], we adopted the weak-coupling 2931 1451 149 model for direct collective inelastic scattering for Cu-65, using 2931 1451 150 Ni-64 as a basis. For the calculation of the cross sections, 2931 1451 151 ECIS96 was used in DWBA mode. We used the following direct 2931 1451 152 transitions for Cu-65 (ground state 3/2- ) : 2931 1451 153 2931 1451 154 Jpi Ex(MeV) Deformation length (Delta) or parameter (Beta) 2931 1451 155 0.5- 0.771 Beta(2)=0.0566 2931 1451 156 2.5- 1.116 Beta(2)=0.0980 2931 1451 157 3.5- 1.481 Beta(2)=0.1132 2931 1451 158 1.5- 1.743 Beta(2)=0.0800 2931 1451 159 1.5- 3.185 Delta(3)=0.3167 2931 1451 160 2.5- 3.435 Delta(3)=0.3879 2931 1451 161 3.5- 3.685 Delta(3)=0.4479 2931 1451 162 4.5- 3.935 Delta(3)=0.5008 2931 1451 163 2931 1451 164 Only one measurement exists for neutron-induced emission spectra 2931 1451 165 above 20 MeV for 65Cu: the 25.7 MeV (n,xn) data by Marcinkowski et2931 1451 166 al (Ma83). Without adjusting any of the level density or pre- 2931 1451 167 equilibrium parameters the GNASH calculation was in good agreement2931 1451 168 with these data (Ko98b). Hence we adopted these parameters for the2931 1451 169 whole energy region. 2931 1451 170 2931 1451 171 **************************************************************** 2931 1451 172 2931 1451 173 REFERENCES 2931 1451 174 2931 1451 175 [Ab93] W. Abfalterer, R.W. Finlay, S.M. Grimes, and V. Mishra, 2931 1451 176 Phys.Rev. C47, 1033 (1993) 2931 1451 177 [Al83] R. Alarcon and J. Rapaport, Nucl.Phys. A458, 502 (1986) 2931 1451 178 [Ar80] E.D. Arthur and P.G. Young, 'Evaluation of Neutron Cross 2931 1451 179 Sections to 40 MeV for 54,56Fe," Proc. Sym. on Neutron Cross 2931 1451 180 Sections from 10 to 50 MeV, 12-14 May 1980, Brookhaven National2931 1451 181 Laboratory [Eds. M. R. Bhat and S. Pearlstein, BNL-NCS- 51245, 2931 1451 182 1980] p. 731. 2931 1451 183 [Be69] F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182, 2931 1451 184 1190 (1969) 2931 1451 185 [Be71] F.D. Becchetti, Jr., and G.W. Greenlees in "Polarization 2931 1451 186 Phenomena in Nuclear Reactions," (Ed: H.H. Barschall and W. 2931 1451 187 Haeberli, The University of Wisconsin Press, 1971) p.682. 2931 1451 188 [Be92] O. Bersillon, "SCAT2 - A Spherical Optical Model Code," 2931 1451 189 in Proc. ICTP Workshop on Computation and Analysis of Nuclear 2931 1451 190 data Relevant to Nuclear Energy and Safety, February-March, 2931 1451 191 1999 Trieste, Italy, to be published in World Scientific Press,2931 1451 192 and Progress Report of the Nuclear Physics Division, Bruyeres- 2931 1451 193 le-Chatel 1977, CEA-N-2037 (1978) p.111 2931 1451 194 [Br50] S. Bratenahl, S. Fernbach, R.H. Hildebrand et al., 2931 1451 195 Phys.Rev. 77, 597 (1950) 2931 1451 196 [Ch93] M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 2931 1451 197 [Ch96] M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J. 2931 1451 198 Koning, "High-Energy Nuclear Data Libraries for Accelerator- 2931 1451 199 Driven Technologies: Calculational Method for Heavy Recoils," 2931 1451 200 Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation 2931 1451 201 Technology and Applications, Kalmar, Sweden, 3-7 June 1996 2931 1451 202 [Ch98] M.B. Chadwick and P.G. Young, "GNASH Calculations of 2931 1451 203 n,p + Cu isotopes and Benchmarking of Results" in APT PROGRESS 2931 1451 204 REPORT: 1 February - 1 March 1998, internal Los Alamos National2931 1451 205 Laboratory memo, 6 Mar.1998 from R.E. MacFarlane to L. Waters. 2931 1451 206 [Ch99] M.B. Chadwick, P G. Young, G. M. Hale, et al., Los Alamos 2931 1451 207 National Laboratory report, LA-UR-99-1222 (1999) 2931 1451 208 [Co67] J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J. 2931 1451 209 Phys. 20, 477 (1967) 2931 1451 210 [De82] J.P. Delaroche, S.M. El-Kadi, P.P. Guss, C.E. Floyd and 2931 1451 211 R.L. Walter, Nucl. Phys. A390, 541 (1982). 2931 1451 212 [El82] S.M. El-Kadi, C.E. Nelson, F.O. Purser et al., Nucl.Phys. 2931 1451 213 A390, 509 (1982) 2931 1451 214 [Fi93] R. W. Finlay, W. P. Abfalterer, G. Fink et al., Phys. Rev 2931 1451 215 C 47, 237 (1993) 2931 1451 216 [Fr01] S.C. Frankle, R.C. Reedy, and P.G. Young, Los ALamos 2931 1451 217 National Laboratory Report, LA-13812 (2001). 2931 1451 218 [Gu86] P. Guenther, D.L. Smith, A.B. Smith, J.F. Whalen, Nucl. 2931 1451 219 Phys. A448, 280 (1986) 2931 1451 220 [Ig75] A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J. 2931 1451 221 Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485 2931 1451 222 (1975) 2931 1451 223 [Ka77] C. Kalbach, Z.Phys.A 283, 401 (1977) 2931 1451 224 [Ka85] C. Kalbach, Los Alamos National Laboratory report 2931 1451 225 LA-10248-MS (1985) 2931 1451 226 [Ka88] C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also 2931 1451 227 C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981) 2931 1451 228 [Ki74] W.E. Kinney, F.G. Perey, report ORNL-4908 (1974) 2931 1451 229 [Ko90] J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990) 2931 1451 230 [Ko97] A.J. Koning, J.J. van Wijk and J.-P. Delaroche, "ECISVIEW:2931 1451 231 A Graphical Interface for ECIS95", Proceedings of the NEA 2931 1451 232 Specialists' Meeting on the Nucleon Nucleus Optical Model up to2931 1451 233 200 MeV, Bruyeres-le-Chatel, November 13-15 1996. Available at 2931 1451 234 http://db.nea.fr/html/science/om200/. 2931 1451 235 [Ko98] A.J. Koning, J.-P. Delaroche and O. Bersillon, "Nuclear 2931 1451 236 Data for Accelerator-Driven Systems: Nuclear models, Experiments2931 1451 237 and Data Libraries", to appear in Mucl. Instr. Meth. A (1998). 2931 1451 238 [Ko98b] A.J. Koning, M.B. Chadwick, and P.G. Young, "ENDF/B-VI 2931 1451 239 neutron and proton datafiles up to 150 MeV for 63Cu and 65Cu", 2931 1451 240 Los Alamos National Laboratory report LAUR- (1998); ECN lab and2931 1451 241 JEFF report (1998). 2931 1451 242 [Lo74] J.M. Lohr and W. Haeberli, Nucl.Phys. A232, 381 (1974) 2931 1451 243 [Ma66] Macfadden and Satchler, Nuc.Phys. 84, 177 (1966) 2931 1451 244 [Ma83] A. Marcinkowski, R.W. Finlay, G. Randers-Pehrson et al., 2931 1451 245 Nucl.Phys. A402, 220 (1983) 2931 1451 246 [Ma88] D.G. Madland, "Recent Results in the Development of a 2931 1451 247 Global Medium-Energy Nucleon-Nucleus Optical-Model Potential, 2931 1451 248 "Proc. OECD/NEANDC Specialist's Mtg. on Preequilibrium Nuclear 2931 1451 249 Reactions, Semmering, Austria, 10-12 Feb. 1988, NEANDC-245 'U' 2931 1451 250 (1988). 2931 1451 251 [Pe63] C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963) 2931 1451 252 [Ra94] J. Raynal, Notes on ECIS94, CEA Saclay Report CEA-N-2772 2931 1451 253 (1994) 2931 1451 254 [Sa60] G.L. Salmon, Nucl.Phys. 21, 15 (1960) 2931 1451 255 [Yo92] P.G. Young, E.D. Arthur, and M.B. Chadwick, report 2931 1451 256 LA-12343-MS (1992) 2931 1451 257 2931 1451 258 **************************************************************** 2931 1451 259 2931 1451 260 ENDF/B-VI MOD 3 Revision, March 1991, ORNL 2931 1451 261 2931 1451 262 MOD 3 changes 2931 1451 263 1) Corrections to MF=6, MT=63 at 17.5 MeV to prevent negative 2931 1451 264 values in the angular distribution. 2931 1451 265 2931 1451 266 **************************************************************** 2931 1451 267 2931 1451 268 * Note there was no MOD 2 released. 2931 1451 269 2931 1451 270 **************************************************************** 2931 1451 271 2931 1451 272 ENDF/B-VI MOD 1 Evaluation, November 1989, D. Hetrick, F.Y. Fu, 2931 1451 273 D. Larson (ORNL) 2931 1451 274 2931 1451 275 This work employed several nuclear model codes including the 2931 1451 276 optical-model code GENOA [1], the Distorted Wave Born 2931 1451 277 Approximation (DWBA) program DWUCK [2], and the Hauser-Feshbach 2931 1451 278 code TNG [3,4]. The TNG code provides energy and angular 2931 1451 279 distributions of particles emitted in the compound and pre- 2931 1451 280 compound reactions, ensures consistency among all reactions, and 2931 1451 281 maintains energy balance. Details pertinent to the contents of 2931 1451 282 this evaluation and extensive comparisons of calculations with 2931 1451 283 experimental data can be found in reference [5]. 2931 1451 284 2931 1451 285 ----- DESCRIPTION OF FILES 2931 1451 286 (MF-MT) 2931 1451 287 1-451 GENERAL INFORMATION, REFERENCES, AND DEFINITIONS. 2931 1451 288 2-151 RESONANCE PARAMETERS WERE TAKEN FROM MUGHABGHAB [6]. POINT2931 1451 289 WISE RECONSTRUCTION COMPARED WITH DATA [7] SHOWED POORER 2931 1451 290 FIT ABOVE 100 KEV, SO THE RESONANCE REGION WAS CUT OFF AT 2931 1451 291 99.5 KEV. REICH-MOORE PARAMETERS ARE GIVEN. AGREEMENT 2931 1451 292 WITH DATA COULD BE IMPROVED WITH ADDITION OF A BACKGROUND 2931 1451 293 FILE IN 3/1, BUT THIS IN GENERAL GIVES TOO LARGE AN 2931 1451 294 AVERAGE CROSS SECTION, WHEN BINNED IN 10 KEV BINS AND 2931 1451 295 COMPARED WITH THE BINNED DATA. THIS IS PROBABLY DUE TO 2931 1451 296 TOO LARGE AN ESTIMATE OF NEUTRON WIDTHS FOR RESONANCES 2931 1451 297 SEEN ONLY IN CAPTURE AND NOT IN TRANSMISSION. 2931 1451 298 NOTE THAT THE FLAG HAS BEEN SET TO ALLOW USER CALCULATION 2931 1451 299 OF THE ANGULAR DISTRIBUTIONS FROM THE R-M RESONANCE 2931 1451 300 PARAMETERS, IF THE USER WANTS ANGULAR DISTRIBUTIONS ON 2931 1451 301 A FINER ENERGY GRID THAN GIVEN IN 4/2. 2931 1451 302 3-16 (N,2N) CROSS SECTIONS WERE TAKEN FROM THE GLUCS [12] 2931 1451 303 CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL- 2931 1451 304 TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2931 1451 305 SECTIONS [13]. 2931 1451 306 33-16 (N,2N) COVARIANCES WERE TAKEN FROM THE GLUCS [12] 2931 1451 307 CALCULATION IN WHICH THIS REACTION WAS STUDIED SIMUL- 2931 1451 308 TANEOUSLY WITH 12 OTHER DOSIMETRY REACTION CROSS 2931 1451 309 SECTIONS [13]. 2931 1451 310 **************************************************************** 2931 1451 311 2931 1451 312 REFERENCES: 2931 1451 313 2931 1451 314 [1] F.G. Perey, computer code GENOA, ORNL, unpublished (1967) 2931 1451 315 [2] P.D. Kunz, "Distorted Wave Code DWUCK72," Univ. of 2931 1451 316 Colorado, unpublished (1972) 2931 1451 317 [3] C.Y. Fu, report ORNL/TM-7042 (1980); also, C.Y Fu, 2931 1451 318 Symp. on Neutron Cross Sections from 10 to 50 MeV, Upton, NY,2931 1451 319 May 1980, Brookhaven National Lab. report BNL-NCS-51245 2931 1451 320 (1980) p.675 2931 1451 321 [4] K. Shibata and C.Y. Fu, report ORNL/TM-10093 (1986) 2931 1451 322 [5] D.M. Hetrick, C.Y. Fu, and D.C. Larson, Oak Ridge report 2931 1451 323 ORNL/TM-9083 [ENDF-337] (1984) 2931 1451 324 [6] S.F. Mughabghab, M. Divadeenam, and N.E. Holden, "Neutron 2931 1451 325 Cross Sections, Vol. 1, Neutron Resonance Parameters and 2931 1451 326 Thermal Cross Sections, Part A, Z=1-60," (Academic Press, 2931 1451 327 1981) 2931 1451 328 [7] M.S. Pandey, J.B. Garg and J.A. Harvey, Phys.Rev. C 15, 600 2931 1451 329 (1977), and private communication. 2931 1451 330 [8] F.G. Perey, private communication (1977) 2931 1451 331 [9] D.C. Larson, Symp. on Neutron Cross Sections from 10 to 50 2931 1451 332 MeV, Upton, NY, May 1980, Brookhaven National Lab. report 2931 1451 333 BNL-NCS-51245 (1980) p.277 2931 1451 334 [10] P. Guenther, D.L. Smith, A.B. Smith and J.F. Whalen, Nucl. 2931 1451 335 Phys. A, 448, 280 (1986) [CSISRS data set 12869/002], and 2931 1451 336 W.P. Poenitz and J.F. Whalen, Argonne report ANL/NDM-80 2931 1451 337 (1983) [CSISRS data set 12853] 2931 1451 338 [11] A.I. Dyumin, D.M. Kaminker, G.N. Popova, and V.A. Smolin, 2931 1451 339 Izv.Akad.Nauk SSSR, Ser.Fiz. 36, 852 (1972) 2931 1451 340 [12] D.M. Hetrick and C.Y. Fu, Oak Ridge report ORNL/TM-7341 2931 1451 341 [ENDF-303] (1980) 2931 1451 342 [13] C.Y. Fu and D.M. Hetrick, Proc. Fourth ASTM-Euratom Symp. 2931 1451 343 on Reactor Dosimetry, Gaithersburg, Maryland, March 22-26, 2931 1451 344 1982 (U.S. National Bureau of Standards) p.877 2931 1451 345 [14] CSISRS Library, National Nuclear Data Center, Brookhaven 2931 1451 346 National Laboratory, Upton, N.Y. 11973. 2931 1451 347 [15] V. McLane, C.L. Dunford and P.F. Rose, "Neutron Cross 2931 1451 348 Sections, Vol. 2, Neutron Cross Section Curves" (Academic 2931 1451 349 Press, 1988) 2931 1451 350 [16] S.M. Qaim and G. Stoecklin, Nucl.Phys. A257, 233 (1976) 2931 1451 351 [17] S.M. Qaim, Radiochimica Acta, 25, 13 (1978) 2931 1451 352 [18] M.G. Delfini, J. Kopecky, R.E. Chrien et al., Nucl.Phys. 2931 1451 353 A404, 250 (1983) 2931 1451 354 **************************************************************** 2931 1451 355 The Q values and threshold energies were updated prior to pro- 2931 1451 356 cessing through the codes to comply with the values obtained 2931 1451 357 using the NNDC calculation program which is based on the 1995 2931 1451 358 Update to the Atomic mass Evaluation. 2931 1451 359 ************************ C O N T E N T S *********************** 2931 1451 360 ***************** Program LINEAR (VERSION 2002-1) ***************2931 1451 361 For All Data Greater than 1.0000E-10 barns in Absolute Value 2931 1451 362 Data Linearized to Within an Accuracy of .100000000 per-cent 2931 1451 363 ***************** Program RECENT (VERSION 2002-1) ***************2931 1451 364 for All Data Greater than 1.0000E-10 barns in Absolute Value 2931 1451 365 Data Linearized to within an Accuracy of .100000000 per-cent 2931 1451 366 ***************** Program SIGMA1 (VERSION 2002-1) ***************2931 1451 367 Data Doppler Broadened to 300.000000 Kelvin 2931 1451 368 for All Data Greater than 1.0000E-10 barns in Absolute Value 2931 1451 369 Data Linearized to Within an Accuracy pf .100000000 per-cent 2931 1451 370 ***************** Program FIXUP (Version 2002-1) ****************2931 1451 371 Corrected ZA/AWR in All Sections-----------------------------Yes 2931 1451 372 Corrected Thresholds-----------------------------------------Yes 2931 1451 373 Extended Cross Sections to 20 MeV----------------------------No 2931 1451 374 Allow Cross Section Deletion---------------------------------No 2931 1451 375 Allow Cross Section Reconstruction---------------------------No 2931 1451 376 Make All Cross Sections Non-Negative-------------------------Yes 2931 1451 377 Delete Energies Not in Ascending Order-----------------------Yes 2931 1451 378 Deleted Duplicate Points-------------------------------------Yes 2931 1451 379 Check for Ascending MAT/MF/MT Order--------------------------Yes 2931 1451 380 Check for Legal MF/MT Numbers--------------------------------Yes 2931 1451 381 Allow Creation of Missing Sections---------------------------No 2931 1451 382 Allow Insertion of Energy Points-----------------------------No 2931 1451 383 Create Uniform Energy Grid-----------------------------------No 2931 1451 384 Delete Section if Cross Section =0 at All Energies-----------Yes 2931 1451 385 ***************** Program GROUPIE (VERSION 2002-1) **************2931 1451 386 Unshielded Group Averages Using 640 Groups 2931 1451 387 Weighting Spectrum: Flat (Constant) Spectrum 2931 1451 388 1 451 392 52931 1451 389 2 151 4 52931 1451 390 3 16 37 12931 1451 391 33 16 62 12931 1451 392 2931 1 099999 2931 0 0 0 2.90650E+4 6.43700E+1 0 0 1 02931 2151 1 2.906500+4 1.000000+0 0 0 1 02931 2151 2 1.000000-5 9.950000+4 0 0 0 02931 2151 3 1.500000+0 6.700000-1 0 0 0 02931 2151 4 2931 2 099999 2931 0 0 0 2.90650E+4 6.43700E+1 0 0 0 02931 3 16 1 -9.91020E+6-9.91020E+6 0 0 1 1012931 3 16 2 101 1 2931 3 16 3 10000000.0 .000366266 10100000.0 .004894626 10200000.0 .0105964472931 3 16 4 10300000.0 .016298268 10400000.0 .022000089 10500000.0 .0367329002931 3 16 5 10600000.0 .060496700 10700000.0 .084260500 10800000.0 .1080243002931 3 16 6 10900000.0 .131788100 11000000.0 .159085000 11100000.0 .1899150002931 3 16 7 11200000.0 .220745000 11300000.0 .251575000 11400000.0 .2824050002931 3 16 8 11500000.0 .314504000 11600000.0 .347872000 11700000.0 .3812400002931 3 16 9 11800000.0 .414608000 11900000.0 .447976000 12000000.0 .4801010002931 3 16 10 12100000.0 .510983000 12200000.0 .541865000 12300000.0 .5727470002931 3 16 11 12400000.0 .603629000 12500000.0 .630458000 12600000.0 .6532340002931 3 16 12 12700000.0 .676010000 12800000.0 .698786000 12900000.0 .7215620002931 3 16 13 13000000.0 .741987000 13100000.0 .760061000 13200000.0 .7781350002931 3 16 14 13300000.0 .796209000 13400000.0 .814283000 13500000.0 .8305190002931 3 16 15 13600000.0 .844917000 13700000.0 .859315000 13800000.0 .8737130002931 3 16 16 13900000.0 .888111000 14000000.0 .900663000 14100000.0 .9113690002931 3 16 17 14200000.0 .922075000 14300000.0 .932781000 14400000.0 .9434870002931 3 16 18 14500000.0 .952211000 14600000.0 .958953000 14700000.0 .9656950002931 3 16 19 14800000.0 .972437000 14900000.0 .979179000 15000000.0 .9856050002931 3 16 20 15100000.0 .991715000 15200000.0 .997825000 15300000.0 1.003935002931 3 16 21 15400000.0 1.01004500 15500000.0 1.01484000 15600000.0 1.018320002931 3 16 22 15700000.0 1.02180000 15800000.0 1.02528000 15900000.0 1.028760002931 3 16 23 16000000.0 1.03373000 16100000.0 1.04019000 16200000.0 1.046650002931 3 16 24 16300000.0 1.05311000 16400000.0 1.05957000 16500000.0 1.063630002931 3 16 25 16600000.0 1.06529000 16700000.0 1.06695000 16800000.0 1.068610002931 3 16 26 16900000.0 1.07027000 17000000.0 1.07157000 17100000.0 1.072510002931 3 16 27 17200000.0 1.07345000 17300000.0 1.07439000 17400000.0 1.075330002931 3 16 28 17500000.0 1.07734000 17600000.0 1.08042000 17700000.0 1.083500002931 3 16 29 17800000.0 1.08658000 17900000.0 1.08966000 18000000.0 1.089750002931 3 16 30 18100000.0 1.08685000 18200000.0 1.08395000 18300000.0 1.081050002931 3 16 31 18400000.0 1.07815000 18500000.0 1.07277167 18600000.0 1.064915002931 3 16 32 18700000.0 1.05705833 18800000.0 1.04920167 18900000.0 1.041345002931 3 16 33 19000000.0 1.03348833 19100000.0 1.02563167 19200000.0 1.017775002931 3 16 34 19300000.0 1.00991833 19400000.0 1.00206167 19500000.0 .9942050002931 3 16 35 19600000.0 .986348333 19700000.0 .978491667 19800000.0 .9706350002931 3 16 36 19900000.0 .962778333 20000000.0 0.0 2931 3 16 37 2931 3 099999 2931 0 0 0 2.90650E+4 6.43700E+1 0 0 0 1293133 16 1 0.0000E+00 0.0000E+00 0 16 0 2293133 16 2 0.0000E+00 0.0000E+00 1 5 300 24293133 16 3 1.0000E-05 1.0000E+07 1.0500E+07 1.1000E+07 1.1500E+07 1.2000E+07293133 16 4 1.2500E+07 1.3000E+07 1.3500E+07 1.4000E+07 1.4500E+07 1.5000E+07293133 16 5 1.5500E+07 1.6000E+07 1.6500E+07 1.7000E+07 1.7500E+07 1.8000E+07293133 16 6 1.8500E+07 1.9000E+07 1.9200E+07 1.9400E+07 1.9600E+07 2.0000E+07293133 16 7 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00293133 16 8 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00293133 16 9 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00293133 16 10 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000E+00 9.2849E-02293133 16 11 -3.9100E-03 9.8124E-04 3.2315E-04 2.5973E-04 1.6943E-04 1.0837E-04293133 16 12 5.2219E-05 4.5076E-05 3.6396E-05 1.4418E-05 2.1146E-05 2.1805E-05293133 16 13 2.1931E-05 2.1318E-05 2.1522E-05 2.2348E-05 2.9580E-05 2.2389E-05293133 16 14 2.5489E-05 2.5389E-05 2.8747E-05 1.1582E-03 9.4407E-05 2.2976E-04293133 16 15 1.6573E-04 1.1914E-04 8.3873E-05 4.3373E-05 3.6631E-05 3.1684E-05293133 16 16 1.1438E-05 1.8007E-05 1.8412E-05 1.8521E-05 1.7912E-05 1.8306E-05293133 16 17 1.8801E-05 2.5446E-05 1.8989E-05 2.1853E-05 2.1701E-05 2.4893E-05293133 16 18 1.0426E-03 4.1291E-04 3.5993E-04 2.4206E-04 1.6476E-04 8.2857E-05293133 16 19 7.0722E-05 6.0039E-05 2.2260E-05 3.4439E-05 3.5246E-05 3.5284E-05293133 16 20 3.4112E-05 3.4653E-05 3.5725E-05 4.7927E-05 3.4885E-05 4.0981E-05293133 16 21 4.0699E-05 4.6995E-05 1.6804E-03 5.1367E-04 3.7077E-04 2.3022E-04293133 16 22 1.0876E-04 9.5181E-05 7.5902E-05 3.0612E-05 4.4708E-05 4.6020E-05293133 16 23 4.5981E-05 4.4753E-05 4.5221E-05 4.6593E-05 6.1095E-05 4.5008E-05293133 16 24 5.2412E-05 5.2045E-05 5.9698E-05 1.5662E-03 4.6015E-04 3.0034E-04293133 16 25 1.3181E-04 1.1846E-04 8.7882E-05 3.8955E-05 5.3423E-05 5.5298E-05293133 16 26 5.5174E-05 5.4093E-05 5.4391E-05 5.6078E-05 7.1784E-05 5.4017E-05293133 16 27 6.1977E-05 6.1527E-05 6.9508E-05 1.6388E-03 3.8739E-04 1.8183E-04293133 16 28 1.6833E-04 1.0963E-04 5.4807E-05 6.8444E-05 7.1184E-05 7.0989E-05293133 16 29 6.9696E-05 7.0329E-05 7.2399E-05 9.1321E-05 7.1362E-05 7.9338E-05293133 16 30 7.8939E-05 8.7688E-05 1.6381E-03 2.7820E-04 2.9569E-04 1.5183E-04293133 16 31 8.4051E-05 9.2687E-05 9.9014E-05 9.7400E-05 9.5524E-05 9.6112E-05293133 16 32 9.8898E-05 1.2500E-04 9.5565E-05 1.0835E-04 1.0739E-04 1.2076E-04293133 16 33 7.2642E-04 1.3861E-04 2.1090E-04 1.3154E-04 1.2590E-04 1.3899E-04293133 16 34 1.3473E-04 1.3248E-04 1.3243E-04 1.3654E-04 1.7246E-04 1.2855E-04293133 16 35 1.4860E-04 1.4753E-04 1.6696E-04 7.4020E-04 2.2844E-04 1.3272E-04293133 16 36 1.3480E-04 1.4740E-04 1.4311E-04 1.4019E-04 1.4023E-04 1.4475E-04293133 16 37 1.8498E-04 1.3536E-04 1.5870E-04 1.5766E-04 1.7994E-04 2.5205E-04293133 16 38 6.7831E-05 1.5774E-04 1.5403E-04 1.5480E-04 1.4692E-04 1.4845E-04293133 16 39 1.5385E-04 2.1458E-04 1.4114E-04 1.7986E-04 1.7866E-04 2.1467E-04293133 16 40 4.9848E-04 7.7213E-05 1.8584E-04 1.5113E-04 1.7339E-04 1.6481E-04293133 16 41 1.6703E-04 1.1799E-04 1.6769E-04 1.2330E-04 1.2252E-04 8.4028E-05293133 16 42 3.4942E-04 1.7432E-04 1.9975E-04 2.0482E-04 1.8271E-04 1.7572E-04293133 16 43 1.6743E-04 1.7022E-04 1.5696E-04 1.5582E-04 1.4329E-04 5.2876E-04293133 16 44 1.9487E-04 2.1525E-04 1.9401E-04 1.9702E-04 1.6642E-04 1.6184E-04293133 16 45 1.5555E-04 1.5458E-04 1.4089E-04 2.9887E-04 1.8750E-04 2.0981E-04293133 16 46 1.8656E-04 1.7332E-04 1.7942E-04 1.6365E-04 1.6255E-04 1.4448E-04293133 16 47 3.3467E-04 2.2613E-04 2.0243E-04 1.7545E-04 1.8875E-04 1.6806E-04293133 16 48 1.6691E-04 1.4225E-04 1.1229E-03-7.4862E-05 3.1309E-04 2.2013E-04293133 16 49 2.7176E-04 2.6985E-04 1.8472E-04 4.9986E-04 1.6697E-04 2.0915E-04293133 16 50 1.7615E-04 1.7495E-04 1.4909E-04 1.3873E-03 1.1714E-04 5.8852E-04293133 16 51 5.8437E-04 3.4575E-04 4.7121E-04 2.5339E-04 2.5176E-04 1.7365E-04293133 16 52 3.0426E-03 1.1867E-03 9.8857E-04 3.0015E-03 9.8197E-04 2.7731E-03293133 16 53 0.0000E+00 0.0000E+00 0 8 48 24293133 16 54 1.0000E-05 0.0000E+00 1.0000E+07 1.8167E-06 1.0500E+07 9.2309E-07293133 16 55 1.1000E+07 5.7456E-06 1.1500E+07 2.8213E-05 1.2000E+07 5.2413E-05293133 16 56 1.2500E+07 8.4423E-05 1.3000E+07 1.1260E-04 1.3500E+07 6.0382E-05293133 16 57 1.4000E+07 7.0182E-05 1.4500E+07 2.6190E-05 1.5000E+07 5.5253E-05293133 16 58 1.5500E+07 4.0563E-05 1.6000E+07 6.4379E-05 1.6500E+07 3.7846E-05293133 16 59 1.7000E+07 4.2917E-05 1.7500E+07 1.4672E-04 1.8000E+07 6.5269E-05293133 16 60 1.8500E+07 1.7207E-04 1.9000E+07 5.8043E-05 1.9200E+07 3.7985E-04293133 16 61 1.9400E+07 3.6370E-04 1.9600E+07 3.0778E-04 2.0000E+07 2.6557E-04293133 16 62 293133 099999 2931 0 0 0 0 0 0 0 3.00640E+4 6.33800E+1 0 0 34 103025 1451 1 0.0 0.0 0 0 0 63025 1451 2 1.00000E+0 2.00000E+7 0 0 10 20023025 1451 3 3.00000E+2 0.0 1 0 42 43025 1451 4 30-Zn- 64 IRK-VIENNA EVAL-APR90 3025 1451 5 DIST-Feb2004 3025 1451 6 ----IRDF-2002 MATERIAL 3025 3025 1451 7 -----INCIDENT NEUTRON DATA 3025 1451 8 ------ENDF-6 FORMAT 3025 1451 9 *****************************************************************3025 1451 10 30-ZN- 64 IRK-VIENNA EVAL-APR90 3025 1451 11 DIST-JUN90 3025 1451 12 IRK-EVAL.NLIB 25 3025 3025 1451 13 *****************************************************************3025 1451 14 The Q value was updated prior to processing through the pre- 3025 1451 15 processing codes to comply with the values obtained using the 3025 1451 16 NNDC calculation program which is based on the 1995 Update to 3025 1451 17 the Atomic mass Evaluation. 3025 1451 18 Threshold value of 0.0 inserted at 1.0e-5 ev for MF/MT=3/103 3025 1451 19 after processing using the pre-processing codes. 3025 1451 20 *****************************************************************3025 1451 21 ***************** Program LINEAR (VERSION 2002-1) ***************3025 1451 22 For All Data Greater than 1.0000E-10 barns in Absolute Value 3025 1451 23 Data Linearized to Within an Accuracy of .100000000 per-cent 3025 1451 24 ***************** Program SIGMA1 (VERSION 2002-1) ***************3025 1451 25 Data Doppler Broadened to 300.000000 Kelvin 3025 1451 26 for All Data Greater than 1.0000E-10 barns in Absolute Value 3025 1451 27 Data Linearized to Within an Accuracy pf .100000000 per-cent 3025 1451 28 ***************** Program FIXUP (Version 2002-1) ****************3025 1451 29 Corrected ZA/AWR in All Sections-----------------------------Yes 3025 1451 30 Corrected Thresholds-----------------------------------------Yes 3025 1451 31 Extended Cross Sections to 20 MeV----------------------------No 3025 1451 32 Allow Cross Section Deletion---------------------------------No 3025 1451 33 Allow Cross Section Reconstruction---------------------------No 3025 1451 34 Make All Cross Sections Non-Negative-------------------------Yes 3025 1451 35 Delete Energies Not in Ascending Order-----------------------Yes 3025 1451 36 Deleted Duplicate Points-------------------------------------Yes 3025 1451 37 Check for Ascending MAT/MF/MT Order--------------------------Yes 3025 1451 38 Check for Legal MF/MT Numbers--------------------------------Yes 3025 1451 39 Allow Creation of Missing Sections---------------------------No 3025 1451 40 Allow Insertion of Energy Points-----------------------------No 3025 1451 41 Create Uniform Energy Grid-----------------------------------No 3025 1451 42 Delete Section if Cross Section =0 at All Energies-----------Yes 3025 1451 43 ***************** Program GROUPIE (VERSION 2002-1) **************3025 1451 44 Unshielded Group Averages Using 640 Groups 3025 1451 45 Weighting Spectrum: Flat (Constant) Spectrum 3025 1451 46 1 451 50 13025 1451 47 2 151 4 13025 1451 48 3 103 72 13025 1451 49 33 103 103 13025 1451 50 3025 1 099999 3025 0 0 0 3.00640E+4 6.33800E+1 0 0 1 03025 2151 1 3.00640E+4 1.00000E+0 0 0 1 03025 2151 2 1.00000E+0 2.00000E+7 0 0 0 03025 2151 3 0.0 6.70000E-1 0 0 0 03025 2151 4 3025 2 099999 3025 0 0 0 3.00640E+4 6.33800E+1 0 0 0 03025 3103 1 2.03600E+5 2.03600E+5 0 0 1 2063025 3103 2 206 1 3025 3103 3 500000.000 2.82258E-6 525000.000 8.46774E-6 550000.000 1.41129E-53025 3103 4 575000.000 1.97581E-5 600000.000 2.59677E-5 630000.000 3.27419E-53025 3103 5 660000.000 3.95161E-5 690000.000 4.62903E-5 720000.000 5.41935E-53025 3103 6 760000.000 6.32258E-5 800000.000 7.22581E-5 840000.000 8.12903E-53025 3103 7 880000.000 9.03226E-5 920000.000 9.93548E-5 960000.000 .0001083873025 3103 8 1000000.00 .000124194 1100000.00 .000182215 1200000.00 .0003133333025 3103 9 1300000.00 .000456625 1400000.00 .000819000 1500000.00 .0012710003025 3103 10 1600000.00 .001886125 1700000.00 .002900000 1800000.00 .0039791253025 3103 11 1900000.00 .006095000 2000000.00 .008635000 2100000.00 .0116475003025 3103 12 2200000.00 .015815000 2300000.00 .020140000 2400000.00 .0267750003025 3103 13 2500000.00 .034355000 2600000.00 .043101625 2700000.00 .0547000003025 3103 14 2800000.00 .066454875 2900000.00 .078801000 3000000.00 .0913890003025 3103 15 3100000.00 .101665125 3200000.00 .106290000 3300000.00 .1107457503025 3103 16 3400000.00 .117132000 3500000.00 .124308000 3600000.00 .1297177503025 3103 17 3700000.00 .130810000 3800000.00 .131846375 3900000.00 .1359710003025 3103 18 4000000.00 .141359000 4100000.00 .146190125 4200000.00 .1496600003025 3103 19 4300000.00 .153109375 4400000.00 .157469000 4500000.00 .1622010003025 3103 20 4600000.00 .166793500 4700000.00 .171045000 4800000.00 .1752502503025 3103 21 4900000.00 .178779000 5000000.00 .182031000 5100000.00 .1846305003025 3103 22 5200000.00 .185635000 5300000.00 .186509125 5400000.00 .1861100003025 3103 23 5500000.00 .185190000 5600000.00 .184270000 5700000.00 .1837981673025 3103 24 5800000.00 .186015333 5900000.00 .188680667 6000000.00 .1913460003025 3103 25 6100000.00 .194011333 6200000.00 .196676667 6300000.00 .1993420003025 3103 26 6400000.00 .202007333 6500000.00 .203560000 6600000.00 .2040000003025 3103 27 6700000.00 .204440000 6800000.00 .204880000 6900000.00 .2053200003025 3103 28 7000000.00 .205760000 7100000.00 .206200000 7200000.00 .2066400003025 3103 29 7300000.00 .207080000 7400000.00 .207520000 7500000.00 .2085860003025 3103 30 7600000.00 .210278000 7700000.00 .211970000 7800000.00 .2136620003025 3103 31 7900000.00 .215354000 8000000.00 .217046000 8100000.00 .2187380003025 3103 32 8200000.00 .220430000 8300000.00 .222122000 8400000.00 .2238140003025 3103 33 8500000.00 .225047333 8600000.00 .225822000 8700000.00 .2265966673025 3103 34 8800000.00 .227371333 8900000.00 .228146000 9000000.00 .2289206673025 3103 35 9100000.00 .229695333 9200000.00 .230470000 9300000.00 .2312446673025 3103 36 9400000.00 .232019333 9500000.00 .232794000 9600000.00 .2335686673025 3103 37 9700000.00 .234343333 9800000.00 .235118000 9900000.00 .2358926673025 3103 38 10000000.0 .237326750 10100000.0 .239420250 10200000.0 .2415137503025 3103 39 10300000.0 .243607250 10400000.0 .245700750 10500000.0 .2477942503025 3103 40 10600000.0 .249887750 10700000.0 .251981250 10800000.0 .2540747503025 3103 41 10900000.0 .256168250 11000000.0 .258261750 11100000.0 .2603552503025 3103 42 11200000.0 .262448750 11300000.0 .264542250 11400000.0 .2666357503025 3103 43 11500000.0 .268729250 11600000.0 .270822750 11700000.0 .2729162503025 3103 44 11800000.0 .275009750 11900000.0 .277103250 12000000.0 .2756823333025 3103 45 12100000.0 .270747000 12200000.0 .265811667 12300000.0 .2608763333025 3103 46 12400000.0 .255941000 12500000.0 .251005667 12600000.0 .2460703333025 3103 47 12700000.0 .241135000 12800000.0 .236199667 12900000.0 .2312643333025 3103 48 13000000.0 .226329000 13100000.0 .221393667 13200000.0 .2164583333025 3103 49 13300000.0 .211523000 13400000.0 .206587667 13500000.0 .2014446673025 3103 50 13600000.0 .196094000 13700000.0 .190743333 13800000.0 .1853926673025 3103 51 13900000.0 .180042000 14000000.0 .174691333 14100000.0 .1693406673025 3103 52 14200000.0 .164595167 14300000.0 .163480667 14400000.0 .1629713333025 3103 53 14500000.0 .162462000 14600000.0 .161564917 14700000.0 .1597200003025 3103 54 14800000.0 .157882450 14900000.0 .157154800 15000000.0 .1568812003025 3103 55 15100000.0 .156607600 15200000.0 .156334000 15300000.0 .1560604003025 3103 56 15400000.0 .155786800 15500000.0 .153846000 15600000.0 .1502380003025 3103 57 15700000.0 .146630000 15800000.0 .143022000 15900000.0 .1394140003025 3103 58 16000000.0 .135806000 16100000.0 .132198000 16200000.0 .1285900003025 3103 59 16300000.0 .124982000 16400000.0 .121374000 16500000.0 .1190220003025 3103 60 16600000.0 .117926000 16700000.0 .116830000 16800000.0 .1157340003025 3103 61 16900000.0 .114638000 17000000.0 .113542000 17100000.0 .1124460003025 3103 62 17200000.0 .111350000 17300000.0 .110254000 17400000.0 .1091580003025 3103 63 17500000.0 .108014000 17600000.0 .106822000 17700000.0 .1056300003025 3103 64 17800000.0 .104438000 17900000.0 .103246000 18000000.0 .1020540003025 3103 65 18100000.0 .100862000 18200000.0 .099670000 18300000.0 .0984780003025 3103 66 18400000.0 .097286000 18500000.0 .096094000 18600000.0 .0949020003025 3103 67 18700000.0 .093710000 18800000.0 .092518000 18900000.0 .0913260003025 3103 68 19000000.0 .090134000 19100000.0 .088942000 19200000.0 .0877500003025 3103 69 19300000.0 .086558000 19400000.0 .085366000 19500000.0 .0841740003025 3103 70 19600000.0 .082982000 19700000.0 .081790000 19800000.0 .0805980003025 3103 71 19900000.0 .079406000 20000000.0 0.0 3025 3103 72 3025 3 099999 3025 0 0 0 3.00640E+4 6.33800E+1 0 0 0 1302533103 1 0.000000+0 0.000000+0 0 103 0 1302533103 2 0.000000+0 0.000000+0 1 5 595 34302533103 3 1.000000-5 5.000000+5 1.250000+6 1.500000+6 1.750000+6 2.000000+6302533103 4 2.250000+6 2.500000+6 2.750000+6 3.000000+6 3.250000+6 3.500000+6302533103 5 3.750000+6 4.000000+6 4.250000+6 4.500000+6 4.750000+6 5.000000+6302533103 6 5.250000+6 5.500000+6 6.000000+6 7.000000+6 8.000000+6 9.000000+6302533103 7 1.100000+7 1.300000+7 1.400000+7 1.450000+7 1.475000+7 1.500000+7302533103 8 1.600000+7 1.700000+7 1.800000+7 2.000000+7 0.000000+0 0.000000+0302533103 9 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0302533103 10 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0302533103 11 0.000000+0 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7.360000-4302533103 88 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0302533103 89 0.000000+0 1.029000-2 5.352000-3 6.455000-3 2.163000-3 4.181000-4302533103 90 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0302533103 91 0.000000+0 8.224000-3 5.780000-3 1.961000-3 3.791000-4 0.000000+0302533103 92 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0302533103 93 1.200000-2 3.205000-3 4.455000-4 0.000000+0 0.000000+0 0.000000+0302533103 94 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.200000-2 1.649000-3302533103 95 7.343000-4 2.925000-4 4.268000-4 1.552000-3 2.783000-3 3.330000-3302533103 96 4.958000-3 1.162000-3 5.634000-4 2.219000-4 4.848000-4 3.209000-4302533103 97 6.358000-4 7.888000-4 1.274000-3 1.183000-3 2.002000-4 5.519000-4302533103 98 5.540000-4 6.012000-4 1.011000-3 2.382000-3 5.401000-4 1.775000-4302533103 99 7.415000-5 1.623000-4 2.084000-4 3.578000-4 1.548000-3 6.073000-4302533103 100 4.197000-4 7.127000-4 1.670000-3 2.862000-3 1.544000-3 2.636000-3302533103 101 6.208000-3 3.103000-3 3.019000-3 6.528000-3 8.573000-3 1.380000-2302533103 102 5.830000-2 302533103 103 302533 099999 3025 0 0 0 0 0 0 0 3.30750E+4 7.42780E+1 0 0 34 103325 1451 1 0.0 0.0 0 0 0 63325 1451 2 1.00000E+0 2.00000E+7 0 0 10 20023325 1451 3 3.00000E+2 0.0 1 0 128 43325 1451 4 33-As- 75 FEI EVAL-Sep94 K.I.Zolotarev 3325 1451 5 DIST-Feb2004 3325 1451 6 ----IRDF-2002 MATERIAL 3325 3325 1451 7 -----INCIDENT NEUTRON DATA 3325 1451 8 ------ENDF-6 FORMAT 3325 1451 9 ***************************************************************** 3325 1451 10 33-AS-75 FEI EVAL-Sep94 K.I.Zolotarev 3325 1451 11 DIST-May03 20030504 3325 1451 12 ----BROND-2 MATERIAL 3325 3325 1451 13 ------Russian Reactor Dosimetry File RRDF-2002 3325 1451 14 ***************************************************************** 3325 1451 15 Authors of evaluation: K.Zolotarev, V.Manokhin, A.Pashchenko 3325 1451 16 ***************************************************************** 3325 1451 17 MF=3 3325 1451 18 MT= 16 - As-75(n,2n)As-74 reaction 3325 1451 19 ------------------------------------- 3325 1451 20 Evaluation was made on the base of experimental data [1-18]. 3325 1451 21 Systematics of (n,2n) excitation functions [19] was taken into 3325 1451 22 account in the energy region 10.5 - 13.0 MeV and 19.0 - 20.0 MeV. 3325 1451 23 Cross section data of E.Paul and R.Clarke [20] were rejected 3325 1451 24 due to inconsistency with main bulk of experimental data and 3325 1451 25 systematics of (n,2n) excitation functions. 3325 1451 26 Statistical analysis of input cross section data was carried 3325 1451 27 out by means of PADE-2 code [21]. Rational function was used as 3325 1451 28 the model function [22]. 3325 1451 29 Integral experimental data for U-235 thermal fission neutron 3325 1451 30 spectrum [23-24] was used for testing evaluated As-75(n,2n)As-74 3325 1451 31 excitation function. The results of testing are given in Table 1. 3325 1451 32 Data for U-235 thermal fission neutron spectrum and Cf-252 3325 1451 33 spontaneous fission neutron spectrum were taken from ref.[25] and 3325 1451 34 [26], respectively. 3325 1451 35 Table 1. 3325 1451 36 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 3325 1451 37 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 3325 1451 38 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 3325 1451 39 U-235 neutron fission ³ 0.30920 ³ 0.304 +- 0.036 [22] 3325 1451 40 ³ ³ 0.311 +- 0.023 [23] 3325 1451 41 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 3325 1451 42 Cf-252 spont. fission ³ 0.61804 ³ 3325 1451 43 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 3325 1451 44 3325 1451 45 MT=33 3325 1451 46 MT= 16 -(n,2n) cross section cov. matrix 3325 1451 47 ---------------------------------------- 3325 1451 48 Uncertainties in the evaluated excitation function for the 3325 1451 49 reaction As-75(n,2n)As-74 are given in the form of relative cova- 3325 1451 50 riance matrix for the 19-neutron energy groups (LB=5). Covariance 3325 1451 51 matrix of uncertainties was calculated simultaneously with 3325 1451 52 recommended cross section data by means of PADE-2 code. 3325 1451 53 Eigenvalues of the 6-th digits relative covariance matrix 3325 1451 54 given in the 33-file are the following: 3325 1451 55 3325 1451 56 2.79921E-09 3.33401E-09 3.76284E-09 3.88107E-09 3325 1451 57 4.19122E-09 4.53852E-09 4.91657E-09 5.52980E-09 3325 1451 58 7.06444E-09 1.01577E-08 1.66580E-08 2.33017E-08 3325 1451 59 1.49576E-07 8.86325E-04 3.07310E-03 3.70656E-03 3325 1451 60 9.37209E-03 1.38518E-02 8.84386E-02 3325 1451 61 3325 1451 62 References : 3325 1451 63 1. R.J.Prestwood, B.P.Bayhurst Phys. Rev., v.121, p.1438, 1961 3325 1451 64 2. B.Granger, M.Longueve Report EANDC(E)-49L, p.83, Oct. 1963 3325 1451 65 3. P.Hille, H.Muenzer Acta Phys. Austriaca, v.23, p.44, May 1966 3325 1451 66 4. J.Csikai, G.Peto Acta Physica Hung., v.23, p.87, May 1967 3325 1451 67 5. M.Bormann et al. Nucl. Phys., v.A115, p.309, July 1968 3325 1451 68 M.Bormann et al. Progress Report EANDC(E)-76, January 1967 3325 1451 69 6. P.R.Prasad et al. Nucl. Phys., v.A138, p.85, November 1969 3325 1451 70 7. P.Venugopala Rao et al. Phys. Rev., v.C3, p.629, Feb. 1971 3325 1451 71 8. M.Wagner, M.Uhl Oesterr. Akad. Wiss., Math.+Naturw. Anzeiger,3325 1451 72 v.108, p.185, November 1971 3325 1451 73 9. S.M.Qaim Nucl. Phys., Sec.A, v.185, p.614, May 1972 3325 1451 74 10. D.V.Viktorov, V.L.Sjablin Jadernaja Fizika, v.15,p.1099, 1972 3325 1451 75 11. C.T.Simpson et al. Inorg. Nucl. Chem., v.35, p.2085,June 1973 3325 1451 76 12. J.Araminowicz,J.Dresler Peport INR-1464, Warsaw,p.14,May 1973 3325 1451 77 13. W.Grochulski et al. Acta Physica Polonica, v.B6,p.139, 1975 3325 1451 78 14. R.A.Sigg Dissertation Abstract, sec.B, v.37,p.2237, Nov.1976 3325 1451 79 15. J.L.Casanova,M.L.Sanchez Anales de Fisica y Quimica (Spain), 3325 1451 80 v.72, n.3, p.186, July 1976 3325 1451 81 16. S.Sothras Dissertation Abstract, sec.B,v.38,p.280, July 1978 3325 1451 82 17. C.Konno, Y.Ikeda, K.Oishi e.a. Report JAERI-1329, Oct. 1993 3325 1451 83 18. I.Birn, S.M.Qaim Nucl. Sci. Eng., v.116, p.125-137, 1994 3325 1451 84 19. V.N.Manokhin VANT, Ser.:Yadernye Konstanty, 1994, No. 1, 3325 1451 85 pp.18-22 3325 1451 86 20. E.B.Paul, R.L.Clarke Can. J. Phys., v.31, p.267, 1953 3325 1451 87 21. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 3325 1451 88 22. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 3325 1451 89 st's Meeting on Evaluation and Processing of Covariance Data, 3325 1451 90 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 3325 1451 91 23. F.Nasyrov, B.D.Sciborskij Atomnaja Energija, v.25, no.5, 3325 1451 92 p.437, November 1968 3325 1451 93 24. E.Steinnes Radiochimica Acta, v.13, p.169, June 1970 3325 1451 94 25. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 3325 1451 95 Library, MAT=9228, MF=5, MT=18, eval. April 1989 3325 1451 96 26. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 3325 1451 97 ***************************************************************** 3325 1451 98 The Q values and threshold energies were updated prior to pro- 3325 1451 99 cessing through the codes to comply with the values obtained 3325 1451 100 using the NNDC calculation program which is based on the 1995 3325 1451 101 Update to the Atomic mass Evaluation. 3325 1451 102 3325 1451 103 File 2 added to the pointwise file containing only the effective 3325 1451 104 scattering radius with no resonance parameters given. 3325 1451 105 Taken from ENDF/B-VI 3325 1451 106 ***************************************************************** 3325 1451 107 ***************** Program LINEAR (VERSION 2002-1) ***************3325 1451 108 For All Data Greater than 1.0000E-10 barns in Absolute Value 3325 1451 109 Data Linearized to Within an Accuracy of .100000000 per-cent 3325 1451 110 ***************** Program SIGMA1 (VERSION 2002-1) ***************3325 1451 111 Data Doppler Broadened to 300.000000 Kelvin 3325 1451 112 for All Data Greater than 1.0000E-10 barns in Absolute Value 3325 1451 113 Data Linearized to Within an Accuracy pf .100000000 per-cent 3325 1451 114 ***************** Program FIXUP (Version 2002-1) ****************3325 1451 115 Corrected ZA/AWR in All Sections-----------------------------Yes 3325 1451 116 Corrected Thresholds-----------------------------------------Yes 3325 1451 117 Extended Cross Sections to 20 MeV----------------------------No 3325 1451 118 Allow Cross Section Deletion---------------------------------No 3325 1451 119 Allow Cross Section Reconstruction---------------------------No 3325 1451 120 Make All Cross Sections Non-Negative-------------------------Yes 3325 1451 121 Delete Energies Not in Ascending Order-----------------------Yes 3325 1451 122 Deleted Duplicate Points-------------------------------------Yes 3325 1451 123 Check for Ascending MAT/MF/MT Order--------------------------Yes 3325 1451 124 Check for Legal MF/MT Numbers--------------------------------Yes 3325 1451 125 Allow Creation of Missing Sections---------------------------No 3325 1451 126 Allow Insertion of Energy Points-----------------------------No 3325 1451 127 Create Uniform Energy Grid-----------------------------------No 3325 1451 128 Delete Section if Cross Section =0 at All Energies-----------Yes 3325 1451 129 ***************** Program GROUPIE (VERSION 2002-1) **************3325 1451 130 Unshielded Group Averages Using 640 Groups 3325 1451 131 Weighting Spectrum: Flat (Constant) Spectrum 3325 1451 132 1 451 136 13325 1451 133 2 151 4 13325 1451 134 3 16 36 13325 1451 135 33 16 42 13325 1451 136 3325 1 099999 3325 0 0 0 3.30750E+4 7.42780E+1 0 0 1 03325 2151 1 3.307500+4 1.000000+0 0 0 1 03325 2151 2 1.000000-5 2.405300+3 0 0 0 03325 2151 3 1.500000+0 0.508630+0 0 0 0 03325 2151 4 3325 2 099999 3325 0 0 0 3.30750E+4 7.42780E+1 0 0 0 03325 3 16 1 -1.02442E+7-1.02442E+7 0 0 1 983325 3 16 2 98 1 3325 3 16 3 10300000.0 6.04805E-6 10400000.0 .000608735 10500000.0 .0052550503325 3 16 4 10600000.0 .014392210 10700000.0 .028074850 10800000.0 .0460172503325 3 16 5 10900000.0 .067864400 11000000.0 .093207450 11100000.0 .1216010003325 3 16 6 11200000.0 .152580000 11300000.0 .185675000 11400000.0 .2204260003325 3 16 7 11500000.0 .256394000 11600000.0 .293214625 11700000.0 .3303858753325 3 16 8 11800000.0 .367557125 11900000.0 .404728375 12000000.0 .4413397503325 3 16 9 12100000.0 .477391250 12200000.0 .512559250 12300000.0 .5468437503325 3 16 10 12400000.0 .580032500 12500000.0 .612125500 12600000.0 .6430125003325 3 16 11 12700000.0 .672693500 12800000.0 .701134000 12900000.0 .7283340003325 3 16 12 13000000.0 .754310750 13100000.0 .779064250 13200000.0 .8026452503325 3 16 13 13300000.0 .825053750 13400000.0 .846360250 13500000.0 .8665647503325 3 16 14 13600000.0 .885747750 13700000.0 .903909250 13800000.0 .9209150003325 3 16 15 13900000.0 .936765000 14000000.0 .952615000 14100000.0 .9672400003325 3 16 16 14200000.0 .980640000 14300000.0 .994040000 14400000.0 1.006381673325 3 16 17 14500000.0 1.01766500 14600000.0 1.02894833 14700000.0 1.039320003325 3 16 18 14800000.0 1.04878000 14900000.0 1.05824000 15000000.0 1.066797503325 3 16 19 15100000.0 1.07445250 15200000.0 1.08210750 15300000.0 1.089762503325 3 16 20 15400000.0 1.09655125 15500000.0 1.10247375 15600000.0 1.108396253325 3 16 21 15700000.0 1.11431875 15800000.0 1.11943900 15900000.0 1.123757003325 3 16 22 16000000.0 1.12807500 16100000.0 1.13239300 16200000.0 1.136711003325 3 16 23 16300000.0 1.14029100 16400000.0 1.14313300 16500000.0 1.145975003325 3 16 24 16600000.0 1.14881700 16700000.0 1.15165900 16800000.0 1.153825003325 3 16 25 16900000.0 1.15531500 17000000.0 1.15680500 17100000.0 1.158295003325 3 16 26 17200000.0 1.15978500 17300000.0 1.16127500 17400000.0 1.162227503325 3 16 27 17500000.0 1.16264250 17600000.0 1.16305750 17700000.0 1.163472503325 3 16 28 17800000.0 1.16335214 17900000.0 1.16269643 18000000.0 1.162040713325 3 16 29 18100000.0 1.16138500 18200000.0 1.16072929 18300000.0 1.160073573325 3 16 30 18400000.0 1.15941786 18500000.0 1.15810500 18600000.0 1.156135003325 3 16 31 18700000.0 1.15416500 18800000.0 1.15219500 18900000.0 1.150225003325 3 16 32 19000000.0 1.14825500 19100000.0 1.14628500 19200000.0 1.143680003325 3 16 33 19300000.0 1.14044000 19400000.0 1.13720000 19500000.0 1.133960003325 3 16 34 19600000.0 1.13072000 19700000.0 1.12748000 19800000.0 1.123815003325 3 16 35 19900000.0 1.11972500 20000000.0 0.0 3325 3 16 36 3325 3 099999 3325 0 0 0 3.30750E+4 7.42780E+1 0 0 0 1332533 16 1 0.000000+0 0.000000+0 0 16 0 1332533 16 2 0.000000+0 0.000000+0 1 5 231 21332533 16 3 1.000000-5 1.030000+7 1.100000+7 1.150000+7 1.200000+7 1.250000+7332533 16 4 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7 1.550000+7332533 16 5 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7 1.850000+7332533 16 6 1.900000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0332533 16 7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0332533 16 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0332533 16 9 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 4.254080-2332533 16 10 3.081380-2 1.931400-2 1.096210-2 5.721690-3 2.776080-3 1.317510-3332533 16 11 7.548290-4 7.048850-4 9.349630-4 1.310960-3 1.762040-3 2.258760-3332533 16 12 2.801620-3 3.416040-3 4.153410-3 5.097380-3 6.377610-3 8.194880-3332533 16 13 2.503660-2 1.710970-2 1.053310-2 5.933080-3 3.031140-3 1.367320-3332533 16 14 5.445340-4 2.757120-4 3.679670-4 6.958200-4 1.178950-3 1.766310-3332533 16 15 2.425790-3 3.137300-3 3.888340-3 4.670750-3 5.478140-3 6.303300-3332533 16 16 1.310600-2 9.036090-3 5.666770-3 3.182410-3 1.508880-3 4.972090-4332533 16 17 1.198130-6-9.989140-5 9.961100-5 5.258840-4 1.118440-3 1.823060-3332533 16 18 2.584600-3 3.338870-3 4.001790-3 4.453280-3 4.511120-3 6.875300-3332533 16 19 4.696000-3 2.853540-3 1.466050-3 5.292000-4-8.635390-6-2.104130-4332533 16 20 -1.367730-4 1.569210-4 6.190120-4 1.197530-3 1.834630-3 2.458670-3332533 16 21 2.972620-3 3.235810-3 3.033720-3 3.453490-3 2.266810-3 1.294430-3332533 16 22 5.856670-4 1.362000-4-8.005410-5-9.600310-5 5.413750-5 3.357960-4332533 16 23 7.122850-4 1.141420-3 1.569950-3 1.925150-3 2.101170-3 1.936260-3332533 16 24 1.633900-3 1.073760-3 6.359020-4 3.318380-4 1.553350-4 9.295540-5332533 16 25 1.286100-4 2.450480-4 4.234780-4 6.421240-4 8.735030-4 1.080130-3332533 16 26 1.207510-3 1.172240-3 8.558900-4 6.667390-4 5.164420-4 4.069720-4332533 16 27 3.365790-4 3.019830-4 2.993640-4 3.246310-4 3.733290-4 4.402050-4332533 16 28 5.184400-4 5.983440-4 6.651090-4 6.747560-4 6.610070-4 6.283810-4332533 16 29 5.803990-4 5.209130-4 4.541900-4 3.852600-4 3.205060-4 2.685700-4332533 16 30 2.417760-4 2.584150-4 3.465920-4 7.554750-4 7.992770-4 7.962510-4332533 16 31 7.522190-4 6.741240-4 5.700760-4 4.499610-4 3.267030-4 2.183230-4332533 16 32 1.513670-4 1.667090-4 9.133160-4 9.708800-4 9.754440-4 9.318050-4332533 16 33 8.455750-4 7.232610-4 5.729600-4 4.056160-4 2.372370-4 9.272540-5332533 16 34 1.098580-3 1.177030-3 1.204940-3 1.180740-3 1.101790-3 9.636420-4332533 16 35 7.593550-4 4.784870-4 1.055990-4 1.346410-3 1.473240-3 1.545330-3332533 16 36 1.546490-3 1.454020-3 1.234990-3 8.406180-4 1.969000-4 1.715590-3332533 16 37 1.907030-3 2.015320-3 1.996100-3 1.785820-3 1.290460-3 3.668920-4332533 16 38 2.227810-3 2.458970-3 2.534180-3 2.358840-3 1.793140-3 6.238740-4332533 16 39 2.814850-3 2.998200-3 2.888540-3 2.307930-3 9.846970-4 3.293890-3332533 16 40 3.287670-3 2.782370-3 1.476680-3 3.433180-3 3.143590-3 2.141300-3332533 16 41 3.284480-3 3.040640-3 4.268410-3 332533 16 42 332533 099999 3325 0 0 0 0 0 0 0 3.90890E+4 8.81421E+1 0 0 34 103925 1451 1 0.0 0.0 0 0 0 63925 1451 2 1.00000E+0 2.00000E+7 0 0 10 20023925 1451 3 3.00000E+2 0.0 1 0 91 43925 1451 4 39-Y - 89 SRI EVAL-JAN96 N.ODANO (SHIP RES. INST.) 3925 1451 5 DIST-Feb2004 3925 1451 6 ----IRDF-2002 MATERIAL 3925 3925 1451 7 -----INCIDENT NEUTRON DATA 3925 1451 8 ------ENDF-6 FORMAT 3925 1451 9 *************************************************************** 3925 1451 10 39- Y- 89 SRI EVAL-JAN96 N.ODANO (SHIP RES. INST.) 3925 1451 11 DIST-JUL98 3925 1451 12 ----JENDL/D-99 MATERIAL 3925 3925 1451 13 *************************************************************** 3925 1451 14 HISTORY 3925 1451 15 96-01 EVALUATION FOR JENDL DOSIMETRY FILE VERSION 2 WAS MADE BY 3925 1451 16 N.ODANO (SHIP RES. INST.). 3925 1451 17 97-09 COMPILED TO JENDL DOSIMETRY FILE 99. 3925 1451 18 3925 1451 19 ==== POINT-WISE DATA FILE ==== 3925 1451 20 3925 1451 21 3925 1451 22 Y-89 (N,2N) Y-88 (HALF-LIFE = 106.65D) 3925 1451 23 3925 1451 24 MF=1 GENERAL INFORMATION 3925 1451 25 MT=451 DESCRIPTIVE DATA AND DICTIONARY 3925 1451 26 3925 1451 27 MF=2 RESONANCE PARAMETERS 3925 1451 28 MT=151 PARAMETERS 3925 1451 29 ONLY SPIN AND SCATTERING RADIUS ARE GIVEN. 3925 1451 30 3925 1451 31 MF=3 NEUTRON CROSS SECTIONS 3925 1451 32 MT=16 (N,2N) CROSS SECTION 3925 1451 33 EXPERIMENTAL DATA/1-17/ IN NESTOR-2/18/ WERE TAKEN FOR 3925 1451 34 THE EVALUATION USING GMA CODE/19/. 3925 1451 35 3925 1451 36 MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS 3925 1451 37 MT=16 GENERATED USING THE GMA CODE. 3925 1451 38 3925 1451 39 REFERENCES 3925 1451 40 1) O.M.HUDSON, JR.+ : BULLETIN OF THE AMERICAN PHYSICAL SOCIETY, 3925 1451 41 6, 506(E9) (1961). 3925 1451 42 2) R.RIEDER+ : EANDC(E)-49L, 83 (1965). 3925 1451 43 3) D.G.VALLIS : AWRE-O-76/66 (1966). 3925 1451 44 4) J.CSIKAI+ : ACTA PHYSICA HANGARIA, 23, 87 (1967). 3925 1451 45 5) D.S.MATHER+ : AWRE-O-47/69 (1969). 3925 1451 46 6) D.R.NETHAWAY : NUCL. PHYS. A, 190, 635 (1972). 3925 1451 47 7) S.M.QAIM+ : EUR-5182E, 939 (1974). 3925 1451 48 8) B.P.BAYHURST+ : PHYS. REV. C, 12, 451 (1975). 3925 1451 49 9) S.K.GHORAI+ : NUCL. PHYS. A, 266, 53 (1976). 3925 1451 50 10) M.BORMANN+ : Z. PHYS. A, 277, 203 (1976). 3925 1451 51 11) L.R.VEESER+ : PHYS. REV. C, 16, 1792 (1977). 3925 1451 52 12) J.FREHAUT+ : PROC. SYMP. ON NEUTRON CROSS SECTIONS FROM 10-50 3925 1451 53 MEV, UPTON, USA, 12-14, 1980, P.399 (1980). 3925 1451 54 13) J.LAUREC+ : CEA-R-5109 (1981). 3925 1451 55 14) P.RAICS+ : ATOMKI KOZLEMENYEK, 23, 45 (1981). 3925 1451 56 15) L.R.GREENWOOD : ASTM-STP-956, 743 (1987). 3925 1451 57 16) HUANG JIAN-ZHOU+ : INDC(CPR)-16 (189). 3925 1451 58 17) C.KONNO+ : JAERI 1329 (1993). 3925 1451 59 18) T.NAKAGAWA : THE JAERI NUCLEAR DATA CENTER, UNPUBLISHED. 3925 1451 60 19) W.P.POENITZ : PROC. CONF. NUCLEAR DATA EVALUATION METHODS 3925 1451 61 AND PROCEDURES, BROOKHAVEN NATIONAL LAB. 1980, BNL-NCS- 3925 1451 62 51363, P.249 (1981). 3925 1451 63 *************************************************************** 3925 1451 64 The Q values and threshold energies were updated prior to pro- 3925 1451 65 cessing through the codes to comply with the values obtained 3925 1451 66 using the NNDC calculation program which is based on the 1995 3925 1451 67 Update to the Atomic mass Evaluation. 3925 1451 68 *************************************************************** 3925 1451 69 3925 1451 70 ***************** Program LINEAR (VERSION 2002-1) ***************3925 1451 71 For All Data Greater than 1.0000E-10 barns in Absolute Value 3925 1451 72 Data Linearized to Within an Accuracy of .100000000 per-cent 3925 1451 73 ***************** Program SIGMA1 (VERSION 2002-1) ***************3925 1451 74 Data Doppler Broadened to 300.000000 Kelvin 3925 1451 75 for All Data Greater than 1.0000E-10 barns in Absolute Value 3925 1451 76 Data Linearized to Within an Accuracy pf .100000000 per-cent 3925 1451 77 ***************** Program FIXUP (Version 2002-1) ****************3925 1451 78 Corrected ZA/AWR in All Sections-----------------------------Yes 3925 1451 79 Corrected Thresholds-----------------------------------------Yes 3925 1451 80 Extended Cross Sections to 20 MeV----------------------------No 3925 1451 81 Allow Cross Section Deletion---------------------------------No 3925 1451 82 Allow Cross Section Reconstruction---------------------------No 3925 1451 83 Make All Cross Sections Non-Negative-------------------------Yes 3925 1451 84 Delete Energies Not in Ascending Order-----------------------Yes 3925 1451 85 Deleted Duplicate Points-------------------------------------Yes 3925 1451 86 Check for Ascending MAT/MF/MT Order--------------------------Yes 3925 1451 87 Check for Legal MF/MT Numbers--------------------------------Yes 3925 1451 88 Allow Creation of Missing Sections---------------------------No 3925 1451 89 Allow Insertion of Energy Points-----------------------------No 3925 1451 90 Create Uniform Energy Grid-----------------------------------No 3925 1451 91 Delete Section if Cross Section =0 at All Energies-----------Yes 3925 1451 92 ***************** Program GROUPIE (VERSION 2002-1) **************3925 1451 93 Unshielded Group Averages Using 640 Groups 3925 1451 94 Weighting Spectrum: Flat (Constant) Spectrum 3925 1451 95 1 451 99 03925 1451 96 2 151 4 03925 1451 97 3 16 32 03925 1451 98 33 16 23 03925 1451 99 3925 1 099999 3925 0 0 0 3.90890E+4 8.81421E+1 0 0 1 03925 2151 1 3.90890E+4 1.00000E+0 0 0 1 03925 2151 2 0.0 0.0 0 0 0 03925 2151 3 5.00000E-1 6.70000E-1 0 0 0 03925 2151 4 3925 2 099999 3925 0 0 0 3.90890E+4 8.81421E+1 0 0 0 03925 3 16 1 -1.14760E+7-1.14760E+7 0 0 1 853925 3 16 2 85 1 3925 3 16 3 11600000.0 .008985210 11700000.0 .030140246 11800000.0 .0559560003925 3 16 4 11900000.0 .082542000 12000000.0 .109128000 12100000.0 .1357140003925 3 16 5 12200000.0 .163801750 12300000.0 .200900000 12400000.0 .2395000003925 3 16 6 12500000.0 .278100000 12600000.0 .316700000 12700000.0 .3553000003925 3 16 7 12800000.0 .393900000 12900000.0 .432500000 13000000.0 .4743200003925 3 16 8 13100000.0 .519360000 13200000.0 .564400000 13300000.0 .6094400003925 3 16 9 13400000.0 .654480000 13500000.0 .692660000 13600000.0 .7239800003925 3 16 10 13700000.0 .755300000 13800000.0 .786620000 13900000.0 .8179400003925 3 16 11 14000000.0 .847270000 14100000.0 .874610000 14200000.0 .9019500003925 3 16 12 14300000.0 .929290000 14400000.0 .956630000 14500000.0 .9775700003925 3 16 13 14600000.0 .992110000 14700000.0 1.00665000 14800000.0 1.021190003925 3 16 14 14900000.0 1.03573000 15000000.0 1.04830000 15100000.0 1.058900003925 3 16 15 15200000.0 1.06950000 15300000.0 1.08010000 15400000.0 1.090700003925 3 16 16 15500000.0 1.10190000 15600000.0 1.11370000 15700000.0 1.125500003925 3 16 17 15800000.0 1.13730000 15900000.0 1.14910000 16000000.0 1.158350003925 3 16 18 16100000.0 1.16505000 16200000.0 1.17175000 16300000.0 1.178450003925 3 16 19 16400000.0 1.18515000 16500000.0 1.19185000 16600000.0 1.198550003925 3 16 20 16700000.0 1.20525000 16800000.0 1.21195000 16900000.0 1.218650003925 3 16 21 17000000.0 1.22130000 17100000.0 1.21990000 17200000.0 1.218500003925 3 16 22 17300000.0 1.21710000 17400000.0 1.21570000 17500000.0 1.214300003925 3 16 23 17600000.0 1.21290000 17700000.0 1.21150000 17800000.0 1.210100003925 3 16 24 17900000.0 1.20870000 18000000.0 1.20790000 18100000.0 1.207700003925 3 16 25 18200000.0 1.20750000 18300000.0 1.20730000 18400000.0 1.207100003925 3 16 26 18500000.0 1.20690000 18600000.0 1.20670000 18700000.0 1.206500003925 3 16 27 18800000.0 1.20630000 18900000.0 1.20610000 19000000.0 1.205900003925 3 16 28 19100000.0 1.20570000 19200000.0 1.20550000 19300000.0 1.205300003925 3 16 29 19400000.0 1.20510000 19500000.0 1.20490000 19600000.0 1.204700003925 3 16 30 19700000.0 1.20450000 19800000.0 1.20430000 19900000.0 1.204100003925 3 16 31 20000000.0 0.0 3925 3 16 32 3925 3 099999 3925 0 0 0 3.90890E+4 8.81421E+1 0 0 0 1392533 16 1 0.000000+0 0.000000+0 0 16 0 1392533 16 2 0.000000+0 0.000000+0 1 5 120 15392533 16 3 1.000000-5 1.160000+7 1.167980+7 1.200000+7 1.262500+7 1.325000+7392533 16 4 1.375000+7 1.425000+7 1.475000+7 1.525000+7 1.575000+7 1.650000+7392533 16 5 1.750000+7 1.900000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0392533 16 6 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0392533 16 7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 2.614690-2392533 16 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0392533 16 9 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0392533 16 10 2.614690-2 2.351680-4 2.278630-4 9.042900-5 9.259900-5 8.525630-5392533 16 11 8.299090-5 7.707100-5 4.918100-5 5.777220-5 5.017710-5 4.334530-5392533 16 12 2.515020-3 1.482340-4 7.056710-5 6.952990-5 6.360270-5 6.802450-5392533 16 13 5.867100-5 4.991930-5 5.263290-5 5.027740-5 3.826150-5 1.181300-3392533 16 14 1.041660-4 9.116020-5 1.023620-4 8.883020-5 1.176510-4 7.032450-5392533 16 15 8.058560-5 7.055540-5 5.527930-5 3.254420-4 1.149980-4 1.205660-4392533 16 16 1.177200-4 1.281940-4 7.931010-5 9.144310-5 8.052500-5 6.472540-5392533 16 17 2.271050-4 1.022200-4 1.102170-4 9.664150-5 6.583630-5 8.648390-5392533 16 18 7.769970-5 7.830730-5 2.030630-4 1.065710-4 1.315170-4 8.313670-5392533 16 19 9.450610-5 8.571720-5 6.963890-5 3.359890-4 1.032510-4 6.690090-5392533 16 20 7.735940-5 6.913090-5 5.707270-5 1.877490-3 1.006380-4 1.132040-4392533 16 21 9.722290-5 7.147720-5 7.645230-4 9.817050-5 1.108810-4 1.009150-4392533 16 22 4.986290-4 1.343200-4 1.551700-4 5.697770-4 1.678400-4 4.251840-4392533 16 23 392533 099999 3925 0 0 0 0 0 0 0 4.00900E+4 8.91324E+1 0 0 34 104025 1451 1 0.0 0.0 0 0 0 64025 1451 2 1.00000E+0 2.00000E+7 0 0 10 20024025 1451 3 3.00000E+2 0.0 1 0 40 44025 1451 4 40-Zr- 90 IRK-VIENNA EVAL-APR90 4025 1451 5 DIST-Feb2004 4025 1451 6 ----IRDF-2002 MATERIAL 4025 4025 1451 7 -----INCIDENT NEUTRON DATA 4025 1451 8 ------ENDF-6 FORMAT 4025 1451 9 *****************************************************************4025 1451 10 40-ZR- 90 IRK-VIENNA EVAL-APR90 4025 1451 11 DIST-JUN90 4025 1451 12 IRK-EVAL.NLIB 25 4025 4025 1451 13 *****************************************************************4025 1451 14 The Q values and threshold energies were updated prior to pro- 4025 1451 15 cessing through the codes to comply with the values obtained 4025 1451 16 using the NNDC calculation program which is based on the 1995 4025 1451 17 Update to the Atomic mass Evaluation. 4025 1451 18 *****************************************************************4025 1451 19 ***************** Program LINEAR (VERSION 2002-1) ***************4025 1451 20 For All Data Greater than 1.0000E-10 barns in Absolute Value 4025 1451 21 Data Linearized to Within an Accuracy of .100000000 per-cent 4025 1451 22 ***************** Program SIGMA1 (VERSION 2002-1) ***************4025 1451 23 Data Doppler Broadened to 300.000000 Kelvin 4025 1451 24 for All Data Greater than 1.0000E-10 barns in Absolute Value 4025 1451 25 Data Linearized to Within an Accuracy pf .100000000 per-cent 4025 1451 26 ***************** Program FIXUP (Version 2002-1) ****************4025 1451 27 Corrected ZA/AWR in All Sections-----------------------------Yes 4025 1451 28 Corrected Thresholds-----------------------------------------Yes 4025 1451 29 Extended Cross Sections to 20 MeV----------------------------No 4025 1451 30 Allow Cross Section Deletion---------------------------------No 4025 1451 31 Allow Cross Section Reconstruction---------------------------No 4025 1451 32 Make All Cross Sections Non-Negative-------------------------Yes 4025 1451 33 Delete Energies Not in Ascending Order-----------------------Yes 4025 1451 34 Deleted Duplicate Points-------------------------------------Yes 4025 1451 35 Check for Ascending MAT/MF/MT Order--------------------------Yes 4025 1451 36 Check for Legal MF/MT Numbers--------------------------------Yes 4025 1451 37 Allow Creation of Missing Sections---------------------------No 4025 1451 38 Allow Insertion of Energy Points-----------------------------No 4025 1451 39 Create Uniform Energy Grid-----------------------------------No 4025 1451 40 Delete Section if Cross Section =0 at All Energies-----------Yes 4025 1451 41 ***************** Program GROUPIE (VERSION 2002-1) **************4025 1451 42 Unshielded Group Averages Using 640 Groups 4025 1451 43 Weighting Spectrum: Flat (Constant) Spectrum 4025 1451 44 1 451 48 14025 1451 45 2 151 4 14025 1451 46 3 16 30 14025 1451 47 33 16 46 14025 1451 48 4025 1 099999 4025 0 0 0 4.00900E+4 8.91324E+1 0 0 1 04025 2151 1 4.00900E+4 1.00000E+0 0 0 1 04025 2151 2 1.00000E+0 2.00000E+7 0 0 0 04025 2151 3 0.0 7.10000E-1 0 0 0 04025 2151 4 4025 2 099999 4025 0 0 0 4.00900E+4 8.91324E+1 0 0 0 04025 3 16 1 -1.19700E+7-1.19700E+7 0 0 1 804025 3 16 2 80 1 4025 3 16 3 12100000.0 .005470824 12200000.0 .017333776 12300000.0 .0342925004025 3 16 4 12400000.0 .056337500 12500000.0 .081985000 12600000.0 .1112350004025 3 16 5 12700000.0 .144872500 12800000.0 .182897500 12900000.0 .2225200004025 3 16 6 13000000.0 .263740000 13100000.0 .303092500 13200000.0 .3405775004025 3 16 7 13300000.0 .376913750 13400000.0 .412101250 13500000.0 .4472887504025 3 16 8 13600000.0 .482476250 13700000.0 .516977500 13800000.0 .5507925004025 3 16 9 13900000.0 .582510000 14000000.0 .612130000 14100000.0 .6424475004025 3 16 10 14200000.0 .673462500 14300000.0 .703267500 14400000.0 .7318625004025 3 16 11 14500000.0 .757600000 14600000.0 .780480000 14700000.0 .8009275004025 3 16 12 14800000.0 .818942500 14900000.0 .837701429 15000000.0 .8572042864025 3 16 13 15100000.0 .876707143 15200000.0 .895960893 15300000.0 .9137200004025 3 16 14 15400000.0 .931230000 15500000.0 .948740000 15600000.0 .9662500004025 3 16 15 15700000.0 .983260583 15800000.0 .997274667 15900000.0 1.010789334025 3 16 16 16000000.0 1.02430400 16100000.0 1.03781867 16200000.0 1.051333334025 3 16 17 16300000.0 1.06484800 16400000.0 1.07836267 16500000.0 1.088679004025 3 16 18 16600000.0 1.09579700 16700000.0 1.10291500 16800000.0 1.110033004025 3 16 19 16900000.0 1.11715100 17000000.0 1.12426900 17100000.0 1.131387004025 3 16 20 17200000.0 1.13850500 17300000.0 1.14562300 17400000.0 1.152741004025 3 16 21 17500000.0 1.15859150 17600000.0 1.16317450 17700000.0 1.167757504025 3 16 22 17800000.0 1.17234050 17900000.0 1.17692350 18000000.0 1.181506504025 3 16 23 18100000.0 1.18608950 18200000.0 1.19067250 18300000.0 1.195255504025 3 16 24 18400000.0 1.19983850 18500000.0 1.20193100 18600000.0 1.201533004025 3 16 25 18700000.0 1.20113500 18800000.0 1.20073700 18900000.0 1.200339004025 3 16 26 19000000.0 1.19994100 19100000.0 1.19954300 19200000.0 1.199145004025 3 16 27 19300000.0 1.19874700 19400000.0 1.19834900 19500000.0 1.197951004025 3 16 28 19600000.0 1.19755300 19700000.0 1.19715500 19800000.0 1.196757004025 3 16 29 19900000.0 1.19635900 20000000.0 0.0 4025 3 16 30 4025 3 099999 4025 0 0 0 4.00900E+4 8.91324E+1 0 0 0 1402533 16 1 0.000000+0 0.000000+0 0 16 0 1402533 16 2 0.000000+0 0.000000+0 1 5 253 22402533 16 3 1.000000-5 1.210000+7 1.240000+7 1.260000+7 1.280000+7 1.300000+7402533 16 4 1.320000+7 1.340000+7 1.360000+7 1.380000+7 1.400000+7 1.420000+7402533 16 5 1.440000+7 1.460000+7 1.480000+7 1.500000+7 1.550000+7 1.600000+7402533 16 6 1.700000+7 1.800000+7 1.900000+7 2.000000+7 0.000000+0 0.000000+0402533 16 7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0402533 16 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0402533 16 9 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0402533 16 10 0.000000+0 1.159000-2 4.915000-3 2.591000-3 2.281000-3 3.480000-3402533 16 11 6.788000-4 1.381000-4 1.958000-4 5.803000-5 5.767000-5 2.137000-4402533 16 12 6.431000-5 5.829000-5 5.247000-5 1.721000-3 1.796000-3 1.230000-3402533 16 13 1.225000-3 1.489000-3 1.297000-3 4.873000-3 2.573000-3 2.266000-3402533 16 14 1.893000-3 6.741000-4 1.371000-4 1.066000-4 5.763000-5 5.727000-5402533 16 15 1.392000-4 6.387000-5 5.789000-5 5.211000-5 1.130000-3 1.177000-3402533 16 16 7.987000-4 7.954000-4 9.667000-4 7.058000-4 8.925000-3 4.716000-3402533 16 17 0.000000+0 1.403000-3 2.854000-4 0.000000+0 1.199000-4 1.192000-4402533 16 18 1.055000-4 1.329000-4 1.205000-4 1.085000-4 8.939000-4 9.207000-4402533 16 19 5.961000-4 5.941000-4 7.214000-4 0.000000+0 6.922000-3 0.000000+0402533 16 20 1.236000-3 2.513000-4 0.000000+0 1.056000-4 1.050000-4 9.288000-5402533 16 21 1.171000-4 1.061000-4 9.551000-5 7.873000-4 8.108000-4 5.250000-4402533 16 22 5.232000-4 6.353000-4 0.000000+0 2.527000-3 0.000000+0 0.000000+0402533 16 23 1.010000-4 0.000000+0 0.000000+0 8.392000-5 0.000000+0 0.000000+0402533 16 24 0.000000+0 6.644000-4 6.964000-4 4.856000-4 4.834000-4 5.878000-4402533 16 25 6.691000-4 1.279000-3 1.354000-4 3.023000-5 5.849000-5 3.123000-5402533 16 26 5.003000-5 5.378000-5 5.446000-5 5.173000-5 2.342000-4 2.412000-4402533 16 27 1.562000-4 1.557000-4 1.890000-4 0.000000+0 3.077000-4 8.119000-5402533 16 28 7.922000-5 7.440000-5 7.199000-5 5.936000-5 7.322000-5 7.150000-5402533 16 29 5.648000-5 5.854000-5 3.739000-5 3.784000-5 4.510000-5 0.000000+0402533 16 30 1.171000-4 4.644000-5 4.378000-5 4.618000-5 3.271000-5 4.260000-5402533 16 31 4.095000-5 3.739000-5 3.920000-5 2.733000-5 2.721000-5 3.308000-5402533 16 32 3.766000-5 9.605000-5 4.489000-5 4.244000-5 3.238000-5 4.414000-5402533 16 33 3.926000-5 2.002000-5 2.062000-5 1.335000-5 1.331000-5 1.616000-5402533 16 34 0.000000+0 1.114000-4 4.042000-5 3.070000-5 4.347000-5 3.718000-5402533 16 35 1.990000-5 2.050000-5 2.254000-5 2.203000-5 2.717000-5 1.259000-5402533 16 36 8.344000-5 2.932000-5 3.968000-5 3.557000-5 5.268000-5 5.499000-5402533 16 37 3.700000-5 3.711000-5 4.471000-5 3.129000-5 6.920000-5 3.915000-5402533 16 38 3.852000-5 2.559000-5 2.649000-5 1.696000-5 1.712000-5 2.046000-5402533 16 39 0.000000+0 1.115000-4 4.736000-5 2.011000-5 2.072000-5 1.341000-5402533 16 40 1.337000-5 1.623000-5 0.000000+0 1.118000-4 2.243000-5 2.328000-5402533 16 41 1.482000-5 1.505000-5 1.786000-5 0.000000+0 6.550000-4 4.446000-4402533 16 42 2.989000-4 2.997000-4 3.613000-4 2.477000-4 6.887000-4 3.120000-4402533 16 43 3.128000-4 3.770000-4 2.596000-4 4.461000-4 3.007000-4 2.902000-4402533 16 44 3.123000-4 4.453000-4 2.888000-4 3.121000-4 5.414000-4 2.681000-4402533 16 45 6.335000-4 402533 16 46 402533 099999 4025 0 0 0 0 0 0 0 4.10930E+4 9.21083E+1 0 0 34 104125 1451 1 0.0 0.0 0 0 0 64125 1451 2 1.00000E+0 2.00000E+7 0 0 10 20024125 1451 3 3.00000E+2 0.0 1 0 408 104125 1451 4 41-Nb- 93 FEI/LANL EVAL-Aug96 Zolotarev et al.,Chadwick et al. 4125 1451 5 DIST-Feb2004 4125 1451 6 ----IRDF-2002 MATERIAL 4125 4125 1451 7 -----INCIDENT NEUTRON DATA 4125 1451 8 ------ENDF-6 FORMAT 4125 1451 9 ***************************************************************** 4125 1451 10 -----RUSSIAN DOSIMETRY FILE RRDF-98 Rev.2 4125 1451 11 ----- MF=3 MT=16 ----- 4125 1451 12 ----- MF=3 MT=51 ----- 4125 1451 13 For the IRDF-2002 file these two reactions were converted at 4125 1451 14 IAEA/NDS. 4125 1451 15 The reaction MF/MT=3/16 was converted to MF/MT=10/16 4125 1451 16 The reaction MF/MT=3/51 was converted to MF/MT=10/ 4 4125 1451 17 The corresponding co-variance files were also converted 4125 1451 18 The reaction MF/MT=33/16 was converted to MF/MT=40/16 4125 1451 19 The reaction MF/MT=33/51 was converted to MF/MT=40/ 4 4125 1451 20 ***************************************************************** 4125 1451 21 41-NB-93 FEI EVAL-Aug96 K.Zolotarev and G.Badikov 4125 1451 22 DIST-Aug96 4125 1451 23 ---BROND-3 MATERIAL 4112 4125 1451 24 ----INCIDENT NEUTRON DATA 4125 1451 25 -----ENDF-6 FORMAT 4125 1451 26 AUTHORS OF EVALUATION: K.Zolotarev and S.Badikov 4125 1451 27 ***************************************************************** 4125 1451 28 41-NB- 93 LANL,ANL EVAL-DEC97 M.CHADWICK,P.YOUNG,D.L.SMITH 4125 1451 29 Ch97,Ch99 DIST-AUG99 REV2- 19990803 4125 1451 30 ----ENDF/B-VI MATERIAL 4125 REVISION 2 4125 1451 31 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************4125 1451 32 DOSIMETRY 4125 1451 33 ******************************************************************4125 1451 34 **************************************************************** 4125 1451 35 ******** Start of (N,2N),(N,N') bibliographical component ******* 4125 1451 36 ***************************************************************** 4125 1451 37 4125 1451 38 ----- MF=3 MT=16 ----- 4125 1451 39 4125 1451 40 The evaluation Nb93(n,2n)Nb92m excitation function from 4125 1451 41 9.20 MeV to 20 MeV has been carried out within the framework of 4125 1451 42 generalized least squares method , rational function was used as 4125 1451 43 model function [1].The experimental data base covers the results 4125 1451 44 of measurements [2-37]. 4125 1451 45 Below 9.2 MeV the evaluation has been carried out by 4125 1451 46 means of linear interpolation. Experimental data [38-43] were 4125 1451 47 rejectected due to large discrepancy with the main bulk of the 4125 1451 48 experimental data. 4125 1451 49 The evaluated Nb93(n,2n)Nb92m excitation function averaged 4125 1451 50 on U-235 neutron fission spectrum [44] and Cf-252 spontaneous 4125 1451 51 fission neutron spectrum [45] gives the next values : 4125 1451 52 4125 1451 53 --------------------------------------------------------------- 4125 1451 54 TYPE OF SPECTRUM I ,MB (calc.) I , MB (measured) 4125 1451 55 ----------------------I-----------------I---------------------- 4125 1451 56 I I 4125 1451 57 U-235 neutron fission I 0.4416 I 0.4796+-0.0293 [46] 4125 1451 58 I I 4125 1451 59 ----------------------I-----------------I---------------------- 4125 1451 60 I I 4125 1451 61 CF-252 spont. fission I 0.7701 I 0.749 +- 0.038 [65] 4125 1451 62 I I 4125 1451 63 4125 1451 64 4125 1451 65 ----- MF=3 MT=51 ----- 4125 1451 66 4125 1451 67 Evaluation Nb93(n,n')Nb93m excitation function from 30 keV 4125 1451 68 to 20 MeV was performed by the same method. Evaluation is 4125 1451 69 based on : 4125 1451 70 from 170 keV to 680 keV on theoretical model calculations [47], 4125 1451 71 from 680 keV to 7.9 MeV on the experimental data [48-51] , 4125 1451 72 from 7.9 MeV to 20.0 MeV on theoretical model calculations [47] 4125 1451 73 and experimental data [52,53]. Below 170 keV the evaluation has 4125 1451 74 been carried out by means of linear interpolation. Experimental 4125 1451 75 data [48] and [52] were renormalized , respectively , to the new 4125 1451 76 monitor cross-sections data U235(n,f) [54] and Nb93(n,2n)Nb92m 4125 1451 77 [55]. 4125 1451 78 Re-evaluated experimental data of M.Wagner e.a. [49-50] were 4125 1451 79 taken from ref. [56]. 4125 1451 80 The evaluated Nb93(n,n')Nb93m excitation function averaged 4125 1451 81 on U-235 neutron fission spectrum [44] and Cf-252 spontaneous 4125 1451 82 fission neutron spectrum [45] gives the next values : 4125 1451 83 4125 1451 84 --------------------------------------------------------------- 4125 1451 85 TYPE OF SPECTRUM I ,MB (calc.) I , MB (measured) 4125 1451 86 ----------------------I-----------------I---------------------- 4125 1451 87 I I 150.0 +- 13.0 [57] 4125 1451 88 U-235 neutron fission I 143.47 I 147.0 +- 11.0 [58] 4125 1451 89 I I 146.2 +- 12.5 [59] 4125 1451 90 ----------------------I-----------------I---------------------- 4125 1451 91 I I 144.9 +- 4.8 [60] 4125 1451 92 CF-252 spont. fission I 146.04 I 144.0 +- 5.0 [61] 4125 1451 93 I I 150.6 +- 3.6 [62] 4125 1451 94 4125 1451 95 The integal cross-section data [57 - 59], measured in U-235 4125 1451 96 neutron fission spectrum, were renormalized to the Nb93m half- 4125 1451 97 life (16.1 +- 0.2) Year [63] , new values of K X-rays branching 4125 1451 98 and new monitor cross-section data. 4125 1451 99 The full description of the cross sections data evaluation 4125 1451 100 for the both reactions is given in [64]. 4125 1451 101 4125 1451 102 References : 4125 1451 103 1. S.Badikov, K.Zolotarev Proc. of NEANSC Specialist's. Meeting 4125 1451 104 on Evaluation and Processing of Covariance Data, Oar Ridge , 4125 1451 105 USA, 1992, OECD, Paris, 1993, p.105 4125 1451 106 2. H.Vonach,H.Muenzer J,OSA, v.96, p.120, Mar.1959 4125 1451 107 3. H.A.Tewes e.a. Report UCRL-6028-T, June 1960 4125 1451 108 4. R.J.Prestwood,B.P.Bayhurst Phys.Rev., v.121, p.1438,Mar.1961 4125 1451 109 5. V.L.Glagolev,P.A.Jampol'skij J,JET, v.13, p.520, 1961 4125 1451 110 6. D.G.Vallis Report AWRE-O-76/66, Dec.1966 4125 1451 111 7. L.Husain e.a. Phys.Rev./C, v.1, p.1233, Apr.1970 4125 1451 112 12. A.Paulsen, R.Widera Z.Physik, v.238, p.23, Sep.1970 4125 1451 113 13. M.Bormann e.a. Progress Rep. EANDC(E)-127, p.38, Mar.1970 4125 1451 114 14. S.M.Qaim e.a. Proc. of Conf.on Chemical Nuclear Data, Measur.4125 1451 115 and Applicat., Univ.of Kent, Canterbury, 20-22 Sep.1971,p.121 4125 1451 116 15. D.R.Nethaway Nucl.Phys./A, v.190, p.635, 1972 4125 1451 117 16. E.Holub,N.Cindro J.of Phys./G, v.2, p.405, June 1976 4125 1451 118 17. D.R.Nethaway J. of Inorg. and Nucl.Chem., v.40, p.1285, 1978 4125 1451 119 18. C.G.Hudson e.a. J.Annals of Nucl.Energy, v.5, p.589, 1978 4125 1451 120 19. J.Laurec e.a. Report CEA-R-5109, June 1981 4125 1451 121 20. T.B.Ryves,P.Kolkowski J.Phys.G, v.7, n.4, p.529, 1981 4125 1451 122 21. R.C.Harper,W.L.Alford J,JP/G, v.8, p.153, Jan.1982 4125 1451 123 22. J.Csikai Proc. of Int.Conf. Nuclear Data for Science and 4125 1451 124 Technology, Antverp., 6-10 Sept.1982, p.414 4125 1451 125 23. A.Chiadli,G.Paic Progress Report MOH-5, p.13, 1982 4125 1451 126 24. S.Daroczy e.a. Conf.,83KIEV, v.3, p.191, Nov.1983 4125 1451 127 25. Y.Ikeda e.a EXFOR 21945, 1984 4125 1451 128 26. Lu Han-Lin e.a. Chinese J.Nucl.Phys., v.6,n.1,p.76, Feb.1984 4125 1451 129 27. I.Garlea e.a. J,RRP, v.29, p.421, 1984 4125 1451 130 28. I.Garlea e.a. Proc.of the 14-th Intern. Symp.on Nucl.Phys., 4125 1451 131 Gaussig,GDR,19-23 Nov.1984, ZFK-562, p.126, 1985 4125 1451 132 29. Y.S.Kim e.a. J.of the Korean Nucl.Soc., v.18, n.2,p.92,1986 4125 1451 133 30. Y.Ikeda e.a. Report JAERI-1312, 1988 4125 1451 134 31. K.Kobayashi,I.Kimura Proc.of Int. Conf.on Nuclear Data for 4125 1451 135 Science and Technology, Mito, Japan 30 May - 3 June 1988 4125 1451 136 32. R.Woelfle e.a. Appl.Radiat.and Isotopes, v.39, p.407, 1988 4125 1451 137 33. Lu Hanlin e.a. Report INDC(CPR)-16, Aug.1989 4125 1451 138 34. Wang Xiuyuan e.a. EXFOR 30935, 1989 4125 1451 139 35. D.C.Santry,R.D.Werner Can.J.Phys., v.68, p.582, 1990 4125 1451 140 36. Y.Ikeda e.a. Progress Report INDC(JPN)-162/U, p.16, Aug.1992 4125 1451 141 37 D.L.Smith e.a. Proc.Int.Conf. Nuclear Data for Science and 4125 1451 142 Technology, Julich, FRG, 13-17 May 1991. Springer Verlag, 4125 1451 143 Berlin - Heidelberg, 1992, p.282-284 4125 1451 144 38. D.R.Koehler,W.L.Alford Report NP-11667, 1962 4125 1451 145 39. F.Strohal e.a. Nucl.Phys., v.30, p.49, Feb.1962 4125 1451 146 40. G.T.Western e.a. Report AFWL-TR-65-216, p.2, June 1966 4125 1451 147 41. W.D.Lu e.a. Phys.Rev.C, v.1, p.350, Jan.1970 4125 1451 148 42. A.Mannan,S.M.Qaim Phys.Rev./C, v.38, n.2, p.630, Aug.1988 4125 1451 149 43. S.I.Bhuiyan EXFOR 30936, 1989 4125 1451 150 44. L.W.Weston e.a. Evaluated Neutron Data for U-235, ENDF/B-VI 4125 1451 151 Library, MAT=9228, MF=5, MT=18, eval.apr. 1989 4125 1451 152 45. W.Mannhart IAEA-TECDOC-410, p.158, 1987 4125 1451 153 46. K.Kobayashi, T.Kobayashi Progress Report NEANDC(J)-155U, 4125 1451 154 P.52, 1990 4125 1451 155 47. B.Strohmaier Ann. nucl. Energy, v.16, n.9, p.461, 1989 4125 1451 156 48. D.B.Gayther e.a. Rep. AERE-R-12612, MAY 1987 4125 1451 157 49. M.Wagner e.a. J.Annals of Nucl.Energy, v.15, n.7, p.363, 1988 4125 1451 158 50. M.Wagner e.a. Proc.Int.Conf. Nucl.Data for Sci.Tec.,Mito, 4125 1451 159 Japan,1988,p.1049 4125 1451 160 51. M.Wagner e.a. Measurement of the activation cross section for 4125 1451 161 the reaction 93Nb(n,n')93mNb in the neutron energy range 4125 1451 162 6 - 9 MeV. Progress Report 1991 4125 1451 163 52. T.B.Ryves,P.Kolkowski J. of Physics G, v.7, n.4, p.529, 1981 4125 1451 164 53. Y.Ikeda e.a. Progress Report INDC(JPN)-162/U, p.16, Aug.1992 4125 1451 165 54. L.W.Weston e.a. Evaluated Neutron Data for U-235, ENDF/B-VI 4125 1451 166 Library, MAT=9228, MF=5, MT=18, eval.apr. 1989 4125 1451 167 55. M.Wagner e.a. Physics Data Nr.13-5, p.175, Karlsruhe 1990 4125 1451 168 56. M.Wagner e.a. Physics Data Nr.13-5, p.135, Karlsruhe 1990 4125 1451 169 57. K.Sakurai, I.Kondo Nucl.Instr.and Meth., v.187, p.649, 1981 4125 1451 170 58. K.Kobayashi, I.Kimura Rep. NEANDC(J)-61U, p.78, sep.1979 4125 1451 171 59. F.Hegedues Proc. of the 1 st ASTM-EURATOM Symposium on Reac- 4125 1451 172 tor Dosimetry, Petten 1975. EUR 5667e, Part 1, p.757, 1976 4125 1451 173 60. W.G.Alberts e.a. Proc. of the 6 th ASTM-EURATOM Symposium on 4125 1451 174 Reactor Dosimetry, ASTM-STP-1OO1, p.223, 1989 4125 1451 175 61. T.G.Williamson e.a. Proc. of the 6 th ASTM-EURATOM Symposium 4125 1451 176 on Reactor Dosimetry, ASTM-STP-1OO1, p.229, 1989 4125 1451 177 62. J.G.Williams e.a. Proc.of the 6 th ASTM-EURATOM Symposium on 4125 1451 178 Reactor Dosimetry, ASTM-STP-1OO1, p.235, 1989 4125 1451 179 63. U.Schotzig,H.Schrader R.PTB-RA-16/2, 2 ed.,Braunschweig,1986 4125 1451 180 64. S.Badikov, K.Zolotarev Proc.of the 8-th ASTM-EURATOM Sympos. 4125 1451 181 on Reactor Dosimetry, Vail, Colorado, USA,29 Aug.-3 Sep. 1993 4125 1451 182 65. W.Mannhart Private communication, 1990 4125 1451 183 ******************************************************************4125 1451 184 ******** End of (N,2N),(N,N') bibliographical component ******** 4125 1451 185 ***************************************************************** 4125 1451 186 ******************************************************************4125 1451 187 ******** Start of (N,G) bibliographical component ******** 4125 1451 188 ***************************************************************** 4125 1451 189 4125 1451 190 ENDF/B-VI MOD 3 Evaluation, December 1997, M.B. Chadwick and 4125 1451 191 P.G. Young (LANL) 4125 1451 192 4125 1451 193 Los Alamos LA150 Library, produced with FKK/GNASH/GSCAN code 4125 1451 194 in cooperation with ECN Petten. 4125 1451 195 4125 1451 196 This evaluation provides a complete representation of the 4125 1451 197 nuclear data needed for transport, damage, heating, radioactivity,4125 1451 198 and shielding applications over the incident proton energy range 4125 1451 199 from 1 to 150 MeV. The evaluation utilizes MF=6, MT=5 to 4125 1451 200 represent all reaction data. Production cross sections and 4125 1451 201 emission spectra are given for neutrons, protons, deuterons, 4125 1451 202 tritons, alpha particles, gamma rays, and all residual nuclides 4125 1451 203 produced (A>5) in the reaction chains. To summarize, the ENDF 4125 1451 204 sections with non-zero data above are: 4125 1451 205 4125 1451 206 MF=3 MT= 2 Integral of nuclear plus interference components 4125 1451 207 of the elastic scattering cross section 4125 1451 208 4125 1451 209 MT= 5 Sum of binary (p,n') and (p,x) reactions 4125 1451 210 4125 1451 211 MF=6 MT= 2 Elastic (p,p) angular distributions given as 4125 1451 212 ratios of the differential nuclear-plus- 4125 1451 213 interference to the integrated value. 4125 1451 214 4125 1451 215 MT= 5 Production cross sections and energy-angle 4125 1451 216 distributions for emission neutrons, protons, 4125 1451 217 deuterons, and alphas; and angle-integrated 4125 1451 218 spectra for gamma rays and residual nuclei that 4125 1451 219 are stable against particle emission 4125 1451 220 4125 1451 221 The evaluation is based on nuclear model calculations that have 4125 1451 222 been benchmarked to experimental data, especially for n +93Nb and 4125 1451 223 n +93Nb reactions [Ch98]. We use the GNASH code system [Yo92], 4125 1451 224 which utilizes Hauser-Feshbach statistical, preequilibrium and 4125 1451 225 direct-reaction theories. Spherical optical model calculations are4125 1451 226 used to obtain particle transmission coefficients for the Hauser- 4125 1451 227 Feshbach calculations, as well as for the elastic proton angular 4125 1451 228 distributions. 4125 1451 229 Cross sections and spectra for producing individual residual 4125 1451 230 nuclei are included for reactions. The energy-angle-correlations 4125 1451 231 for all outgoing particles are based on Kalbach systematics 4125 1451 232 [Ka88]. 4125 1451 233 A model was developed to calculate the energy distributions of 4125 1451 234 all recoil nuclei in the GNASH calculations [Ch96a]. The recoil 4125 1451 235 energy distributions are represented in the laboratory system in 4125 1451 236 MT=5, MF=6, and are given as isotropic in the lab system. All 4125 1451 237 other data in MT=5,MF=6 are given in the center-of-mass system. 4125 1451 238 This method of representation utilizes the LCT=3 option approved 4125 1451 239 at the November, 1996, CSEWG meeting. 4125 1451 240 Preequilibrium corrections were performed in the course of the 4125 1451 241 GNASH calculations using the exciton model of Kalbach [Ka77, 4125 1451 242 Ka85], validated by comparison with calculations using Feshbach, 4125 1451 243 Kerman, Koonin (FKK) theory [Ch93]. Discrete level data from 4125 1451 244 nuclear data sheets were matched to continuum level densities 4125 1451 245 using the formulation of Ignatyuk et al. [Ig75] and pairing and 4125 1451 246 shell parameters from the Cook [Co67] analysis. Neutron and 4125 1451 247 charged- particle transmission coefficients were obtained from the4125 1451 248 optical potentials, as discussed below. Gamma-ray transmission 4125 1451 249 coefficients were calculated using the Kopecky-Uhl model [Ko90]. 4125 1451 250 4125 1451 251 SPECIFIC INFORMATION CONCERNING THE 93Nb EVALUATION 4125 1451 252 4125 1451 253 The total neutron cross section was obtained from the Finlay[Fi93]4125 1451 254 measurements. 4125 1451 255 4125 1451 256 The following optical potentials were used in the GNASH 4125 1451 257 calculations. For incident neutrons, the Wilmore-Hodgson 4125 1451 258 potential was used below 15 MeV, and the Madland potential [Ma88] 4125 1451 259 was used at higher energies. For incident protons, the 4125 1451 260 Becchetti-Greenlees [Be69] potential was used up to 50 MeV, above 4125 1451 261 which the Madland potential [Ma88] was used. In both cases, the 4125 1451 262 matching energy between the potentials was chosen to result in 4125 1451 263 continuity of the reaction cross section. For protons at 50 MeV 4125 1451 264 the reaction cross section (and transmission coefficients) was 4125 1451 265 renormalized slightly to smoothen the transition between the 4125 1451 266 potentials. The Perey [Pe63] potential was used for indident 4125 1451 267 deuterons. For tritons, the Becchetti-Greenlees [Be71] was used 4125 1451 268 up to 80 MeV, above which the Watanabe potential was used. The 4125 1451 269 Moyen (McFadden Satchler) [Mc66] potential was used for alpha 4125 1451 270 particles over the whole energy range. 4125 1451 271 4125 1451 272 Direct inelastic scattering to low-lying states in Nb93 was 4125 1451 273 determined as follows. Coherent excitation of 2+ and 3- 4125 1451 274 vibrations were assumed to be fragmented over Nb93 states, after 4125 1451 275 coupling these excitations with the 4.5+ core. The magnitudes of 4125 1451 276 the deformation lengths of 2+ and 3- excitations was obtained by 4125 1451 277 fitting values of 34 and 46 mb respectively at 14 MeV, obtained 4125 1451 278 in ref. [Ch93] and accounting for measurements well. This 4125 1451 279 strength was then fragmented over Nb states. For the 3- 4125 1451 280 excitation, the 7 states are in the "continuum" region of the 4125 1451 281 GNASH calculation at approximately 2.5 MeV, with spins 1.5-,2.5-, 4125 1451 282 ..,7.5-. For the 2+, the 5 states (2.5+,3.5+,...6.5+) near 1 MeV 4125 1451 283 were assumed to be those whose inelastic cross section in the 4125 1451 284 existing ENDF <20 MeV file are significant (note that the ENDF 4125 1451 285 file below 20 MeV appears to incorporate inelastic information 4125 1451 286 only up to 5 MeV for many states, after which a value of zero at 4125 1451 287 20 MeV was inserted). 4125 1451 288 4125 1451 289 Experimental data is used to benchmark the calculations. For 4125 1451 290 incident neutrons, experimental neutron emission spectra data 4125 1451 291 exist at 20 and 26 MeV by Marcinkowski [Ma83]. For incident 4125 1451 292 protons, spectra data exist at 14 and 26 MeV by Watanabe et 4125 1451 293 al. [Wa97], and at 65 MeV by Sakai et al [Sa80]. Our evaluation 4125 1451 294 agrees reasonably well with these measurements. 4125 1451 295 4125 1451 296 **************************************************************** 4125 1451 297 4125 1451 298 REFERENCES 4125 1451 299 4125 1451 300 [Be69] F.D. Becchetti, Jr., and G.W. Greenlees, Phys.Rev. 182, 4125 1451 301 1190 (1969) 4125 1451 302 [Be71] F.D. Becchetti, Jr., and G.W. Greenlees in Polarization 4125 1451 303 Phenomena in Nuclear Reactions (Ed: H.H. Barschall and W. 4125 1451 304 Haeberli, The University of Wisconsin Press, 1971) p.682 4125 1451 305 [Ch93] M.B. Chadwick and P.G. Young, Phys.Rev. C 47, 2255 (1993) 4125 1451 306 [Ch96a] M.B. Chadwick, P.G. Young, R.E. MacFarlane, and A.J. 4125 1451 307 Koning, "High-Energy Nuclear Data Libraries for Accelerator- 4125 1451 308 Driven Technologies: Calculational Method for Heavy Recoils," 4125 1451 309 Proc. of 2nd Int. Conf. on Accelerator Driven Transmutation 4125 1451 310 Technology and Applications, Kalmar, Sweden, 3-7 June 1996. 4125 1451 311 [Ch98] M.B. Chadwick and P.G. Young, "Model Calculations of 4125 1451 312 n,p + 93Nb" in APT PROGRESS REPORT: 1 November 1997 - 1 January 4125 1451 313 1998, internal Los Alamos National Laboratory memo 4125 1451 314 January 1998 from R.E. MacFarlane to L. Waters. 4125 1451 315 [Ch99] M.B. Chadwick, P.G. Young, G.M. Hale et al., Los Alamos 4125 1451 316 National Laboratory report, LA-UR-99-1222 (1999) 4125 1451 317 [Co67] J.L. Cook, H. Ferguson, and A.R. DeL Musgrove, Aust.J. 4125 1451 318 Phys. 20, 477 (1967) 4125 1451 319 [Fi93] R. W. Finlay, W. P. Abfalterer, G. Fink et al., Phys. Rev 4125 1451 320 C 47, 237 (1993) 4125 1451 321 [Ig75] A.V. Ignatyuk, G.N. Smirenkin, and A.S. Tishin, Sov.J. 4125 1451 322 Nucl.Phys. 21, 255 (1975); translation of Yad.Fiz. 21, 485 4125 1451 323 (1975) 4125 1451 324 [Ka77] C. Kalbach, Z.Phys.A 283, 401 (1977) 4125 1451 325 [Ka85] C. Kalbach, Los Alamos National Laboratory report 4125 1451 326 LA-10248-MS (1985) 4125 1451 327 [Ka88] C. Kalbach, Phys.Rev.C 37, 2350 (1988); see also 4125 1451 328 C. Kalbach and F. M. Mann, Phys.Rev.C 23, 112 (1981) 4125 1451 329 [Ko90] J. Kopecky and M. Uhl, Phys.Rev.C 41, 1941 (1990) 4125 1451 330 [Lo74] J.M. Lohr and W. Haeberli, Nucl.Phys. A232, 381 (1974) 4125 1451 331 [Ma88] D.G. Madland, "Recent Results in the Development of a 4125 1451 332 Global Medium-Energy Nucleon-Nucleus Optical-Model Potential," 4125 1451 333 Proc. OECD/NEANDC Specialist's Mtg. on Preequilibrium Nuclear 4125 1451 334 Reactions, Semmering, Austria, 10-12 Feb. 1988, NEANDC-245 'U' 4125 1451 335 (1988). 4125 1451 336 [Ma83] A. Marcinkowski, R.W. Finlay, G. Randers-Pehrson et al., 4125 1451 337 Nucl.Phys. A402, 220 (1983) 4125 1451 338 [Mc66] L. McFadden and G. R. Satchler, Nucl. Phys. 84, 177 4125 1451 339 (1966). 4125 1451 340 [Pe63] C.M. Perey and F.G. Perey, Phys.Rev. 132, 755 (1963) 4125 1451 341 [Sa80] H. Sakai, K. Hosono, N. Matsuoka et al., Nucl.Phys. A344, 4125 1451 342 41 (1980) 4125 1451 343 [We96] H.P. Wellisch and D. Axen, Phys.Rev. C 54, 1329(1996) 4125 1451 344 [Wa97] Y. Watanabe, S. Yoshioka, M. Harada et al, Nuclear Data 4125 1451 345 for Science and Technology, Proc. Conf. Trieste, May, 1997, 4125 1451 346 G. Reffo, Ed. (Editrice Compositori, 1997) p.580 4125 1451 347 [Wi64] D. Wilmore and P.E. Hodgson, Nucl.Phys. 55, 673 (1964) 4125 1451 348 [Yo92] P.G. Young, E.D. Arthur, and M.B. Chadwick, report 4125 1451 349 LA-12343-MS (1992) 4125 1451 350 4125 1451 351 ***************************************************************** 4125 1451 352 4125 1451 353 ENDF/B-VI MOD 2 Revision, August 1991, NNDC 4125 1451 354 4125 1451 355 Only the section MOD numbers have been corrected in the 4125 1451 356 directory. 4125 1451 357 4125 1451 358 ***************************************************************** 4125 1451 359 4125 1451 360 ENDF/B-VI MOD 1 Evaluation, March 1990, A.B. Smith, D.L. Smith 4125 1451 361 L.P. Geraldo (ANL), and R. Howerton (LLNL) 4125 1451 362 4125 1451 363 Original evaluation fully documented in Smith et al. [1] 4125 1451 364 4125 1451 365 ---------------------------------------------------------------- 4125 1451 366 REFERENCES 4125 1451 367 4125 1451 368 [1.] A.B. Smith, D.L. Smith and R.J. Howerton, Argonne report 4125 1451 369 ANL/NDM-88 (1985) 4125 1451 370 [2.] D.L. Smith and L.P. Geraldo, Argonne report ANL/NDM-117 4125 1451 371 (1990) 4125 1451 372 4125 1451 373 ******************************************************************4125 1451 374 ******** End of (N,G) bibliographical component ******** 4125 1451 375 ***************************************************************** 4125 1451 376 The Q values and threshold energies were updated prior to pro- 4125 1451 377 cessing through the codes to comply with the values obtained 4125 1451 378 using the NNDC calculation program which is based on the 1995 4125 1451 379 Update to the Atomic mass Evaluation. 4125 1451 380 4125 1451 381 File 2 taken from ENDF/B-VI 4125 1451 382 ************************ C O N T E N T S *********************** 4125 1451 383 4125 1451 384 ***************** Program LINEAR (VERSION 2002-1) ***************4125 1451 385 For All Data Greater than 1.0000E-10 barns in Absolute Value 4125 1451 386 Data Linearized to Within an Accuracy of .100000000 per-cent 4125 1451 387 ***************** Program RECENT (VERSION 2002-1) ***************4125 1451 388 for All Data Greater than 1.0000E-10 barns in Absolute Value 4125 1451 389 Data Linearized to within an Accuracy of .100000000 per-cent 4125 1451 390 ***************** Program SIGMA1 (VERSION 2002-1) ***************4125 1451 391 Data Doppler Broadened to 300.000000 Kelvin 4125 1451 392 for All Data Greater than 1.0000E-10 barns in Absolute Value 4125 1451 393 Data Linearized to Within an Accuracy pf .100000000 per-cent 4125 1451 394 ***************** Program FIXUP (Version 2002-1) ****************4125 1451 395 Corrected ZA/AWR in All Sections-----------------------------Yes 4125 1451 396 Corrected Thresholds-----------------------------------------Yes 4125 1451 397 Extended Cross Sections to 20 MeV----------------------------No 4125 1451 398 Allow Cross Section Deletion---------------------------------No 4125 1451 399 Allow Cross Section Reconstruction---------------------------No 4125 1451 400 Make All Cross Sections Non-Negative-------------------------Yes 4125 1451 401 Delete Energies Not in Ascending Order-----------------------Yes 4125 1451 402 Deleted Duplicate Points-------------------------------------Yes 4125 1451 403 Check for Ascending MAT/MF/MT Order--------------------------Yes 4125 1451 404 Check for Legal MF/MT Numbers--------------------------------Yes 4125 1451 405 Allow Creation of Missing Sections---------------------------No 4125 1451 406 Allow Insertion of Energy Points-----------------------------No 4125 1451 407 Create Uniform Energy Grid-----------------------------------No 4125 1451 408 Delete Section if Cross Section =0 at All Energies-----------Yes 4125 1451 409 ***************** Program GROUPIE (VERSION 2002-1) **************4125 1451 410 Unshielded Group Averages Using 640 Groups 4125 1451 411 Weighting Spectrum: Flat (Constant) Spectrum 4125 1451 412 1 451 422 14125 1451 413 2 151 4 14125 1451 414 3 102 217 14125 1451 415 8 4 2 14125 1451 416 8 16 2 14125 1451 417 10 4 90 14125 1451 418 10 16 41 14125 1451 419 33 102 9 14125 1451 420 40 4 104 14125 1451 421 40 16 39 14125 1451 422 4125 1 099999 4125 0 0 0 4.10930E+4 9.21083E+1 0 0 1 04125 2151 1 4.10930E+4 1.00000E+0 0 0 1 04125 2151 2 1.00000E-5 7.35000E+3 0 0 0 04125 2151 3 4.50000E+0 6.90000E-1 0 0 0 04125 2151 4 4125 2 099999 4125 0 0 0 4.10930E+4 9.21083E+1 0 0 0 04125 3102 1 7.22747E+6 7.22747E+6 0 0 1 6414125 3102 2 641 1 4125 3102 3 .000100000 18.0213007 .000105000 17.6011498 .000110000 17.20200384125 3102 4 .000115000 16.8342057 .000120000 16.4042406 .000127500 15.92720694125 3102 5 .000135000 15.4913843 .000142500 15.0908610 .000150000 14.65737964125 3102 6 .000160000 14.2058190 .000170000 13.7949650 .000180000 13.41782814125 3102 7 .000190000 13.0669303 .000200000 12.7465657 .000210000 12.44515734125 3102 8 .000220000 12.1651963 .000230000 11.9050847 .000240000 11.59977434125 3102 9 .000255000 11.2642942 .000270000 11.0054074 .000280000 10.71720284125 3102 10 .000300000 10.3657565 .000320000 10.0475333 .000340000 9.756094094125 3102 11 .000360000 9.48766816 .000380000 9.24263675 .000400000 8.987176444125 3102 12 .000425000 8.72576587 .000450000 8.48693672 .000475000 8.267962614125 3102 13 .000500000 8.06202911 .000525000 7.87383007 .000550000 7.696048704125 3102 14 .000575000 7.53070746 .000600000 7.36124432 .000630000 7.186911624125 3102 15 .000660000 7.02739729 .000690000 6.87437954 .000720000 6.711564314125 3102 16 .000760000 6.53699045 .000800000 6.37490446 .000840000 6.226363634125 3102 17 .000880000 6.08518481 .000920000 5.95576328 .000960000 5.831819854125 3102 18 .001000000 5.70335266 .001050000 5.56924720 .001100000 5.443531344125 3102 19 .001150000 5.32768145 .001200000 5.19101925 .001275000 5.040835304125 3102 20 .001350000 4.90352207 .001425000 4.77537192 .001500000 4.639231264125 3102 21 .001600000 4.49688317 .001700000 4.36693655 .001800000 4.246754304125 3102 22 .001900000 4.13696432 .002000000 4.03517330 .002100000 3.939658444125 3102 23 .002200000 3.85218040 .002300000 3.76853368 .002400000 3.673114994125 3102 24 .002550000 3.56708007 .002700000 3.48413693 .002800000 3.394028254125 3102 25 .003000000 3.28311047 .003200000 3.18218089 .003400000 3.089889654125 3102 26 .003600000 3.00571857 .003800000 2.92800045 .004000000 2.846995734125 3102 27 .004250000 2.76475415 .004500000 2.68968605 .004750000 2.619526324125 3102 28 .005000000 2.55514441 .005250000 2.49546942 .005500000 2.439062994125 3102 29 .005750000 2.38736888 .006000000 2.33308297 .006300000 2.278836274125 3102 30 .006600000 2.22751075 .006900000 2.17992067 .007200000 2.128184914125 3102 31 .007600000 2.07287015 .008000000 2.02229362 .008400000 1.974836684125 3102 32 .008800000 1.93068724 .009200000 1.88965925 .009600000 1.850618324125 3102 33 .010000000 1.81018797 .010500000 1.76743310 .011000000 1.728284414125 3102 34 .011500000 1.69092729 .012000000 1.64810317 .012750000 1.600693434125 3102 35 .013500000 1.55678814 .014250000 1.51653646 .015000000 1.473662134125 3102 36 .016000001 1.42869772 .017000001 1.38751152 .017999999 1.349987664125 3102 37 .018999999 1.31541292 .020000000 1.28292398 .021000000 1.253192614125 3102 38 .022000000 1.22518032 .023000000 1.19908065 .024000000 1.168416244125 3102 39 .025500000 1.13534352 .027000001 1.10913958 .028000001 1.080676714125 3102 40 .029999999 1.04563566 .032000002 1.01370832 .034000002 .9845640514125 3102 41 .035999998 .957974892 .037999999 .933469317 .039999999 .9079185714125 3102 42 .042500000 .882001933 .045000002 .858314863 .047499999 .8362348154125 3102 43 .050000001 .815903693 .052499998 .797076973 .055000000 .7792451944125 3102 44 .057500001 .762920918 .059999999 .745757234 .063000001 .7285945024125 3102 45 .066000000 .712422526 .068999998 .697360082 .071999997 .6810306544125 3102 46 .075999998 .663545193 .079999998 .647586836 .083999999 .6326033124125 3102 47 .088000000 .618689959 .092000000 .605799476 .096000001 .5934841284125 3102 48 .100000001 .580740692 .104999997 .567238550 .109999999 .5548929024125 3102 49 .115000002 .543099907 .119999997 .529595519 .127499998 .5146728974125 3102 50 .135000005 .500837179 .142499998 .488141351 .150000006 .4746142334125 3102 51 .159999996 .460419968 .170000002 .447512045 .180000007 .4358671434125 3102 52 .189999998 .424682446 .200000003 .414537148 .209999993 .4051062834125 3102 53 .219999999 .396226892 .230000004 .388071824 .239999995 .3784338824125 3102 54 .254999995 .367847380 .270000011 .359748630 .280000001 .3506513334125 3102 55 .300000012 .339586358 .319999993 .329572955 .340000004 .3204548804125 3102 56 .360000014 .312085870 .379999995 .304512822 .400000006 .2963389924125 3102 57 .425000012 .288129965 .449999988 .280647711 .474999994 .2736699674125 3102 58 .500000000 .267249196 .524999976 .261324591 .550000012 .2557032224125 3102 59 .574999988 .250569355 .600000024 .245164713 .629999995 .2397563244125 3102 60 .660000026 .234629738 .689999998 .229956081 .720000029 .2249389214125 3102 61 .759999990 .219377641 .800000012 .214412077 .839999974 .2095743734125 3102 62 .879999995 .205287449 .920000017 .201117331 .959999979 .1972194084125 3102 63 1.00000000 .193144171 1.04999995 .188952517 1.10000002 .1849860024125 3102 64 1.14999998 .181411664 1.20000005 .177069478 1.27499998 .1722936074125 3102 65 1.35000002 .168067132 1.42499995 .163992345 1.50000000 .1598027644125 3102 66 1.60000002 .155279789 1.70000005 .151227126 1.79999995 .1473994314125 3102 67 1.89999998 .143982571 2.00000000 .140708632 2.09999990 .1377903334125 3102 68 2.20000005 .134930277 2.29999995 .132383342 2.40000010 .1293041244125 3102 69 2.54999995 .125954380 2.70000005 .123359271 2.79999995 .1205590374125 3102 70 3.00000000 .117003541 3.20000005 .113818356 3.40000010 .1109472904125 3102 71 3.59999990 .108223278 3.79999995 .105805012 4.00000000 .1031868124125 3102 72 4.25000000 .100605679 4.50000000 .098236392 4.75000000 .0960118974125 3102 73 5.00000000 .093988893 5.25000000 .092041920 5.50000000 .0902770754125 3102 74 5.75000000 .088613386 6.00000000 .086909289 6.30000019 .0851493054125 3102 75 6.59999990 .083560185 6.90000010 .082063272 7.19999981 .0803861634125 3102 76 7.59999990 .078663891 8.00000000 .077001855 8.39999962 .0755136994125 3102 77 8.80000019 .074118084 9.19999981 .072831750 9.60000038 .0715604124125 3102 78 10.0000000 .070273917 10.5000000 .068923387 11.0000000 .0677183804125 3102 79 11.5000000 .066530440 12.0000000 .065212677 12.7500000 .0637066964125 3102 80 13.5000000 .062376173 14.2500000 .061106755 15.0000000 .0598156094125 3102 81 16.0000000 .058434376 17.0000000 .057233258 18.0000000 .0561099024125 3102 82 19.0000000 .055122657 20.0000000 .054279032 21.0000000 .0534859204125 3102 83 22.0000000 .052806123 23.0000000 .052221292 24.0000000 .0516809544125 3102 84 25.5000000 .051321727 27.0000000 .051340932 28.0000000 .0521454644125 3102 85 30.0000000 .056070388 32.0000000 .075305021 34.0000000 2.145136594125 3102 86 36.0000000 1.09955184 38.0000000 .084352271 40.0000000 1.343405374125 3102 87 42.5000000 .413498725 45.0000000 .054923719 47.5000000 .0464178864125 3102 88 50.0000000 .043601210 52.5000000 .042150748 55.0000000 .0412323614125 3102 89 57.5000000 .040589832 60.0000000 .040061649 63.0000000 .0396397084125 3102 90 66.0000000 .039357775 69.0000000 .039188537 72.0000000 .0391362604125 3102 91 76.0000000 .039334796 80.0000000 .040032723 84.0000000 .0421793534125 3102 92 88.0000000 .054697646 92.0000000 1.90564800 96.0000000 .0679057414125 3102 93 100.000000 .105279263 105.000000 1.74244681 110.000000 .0804239034125 3102 94 115.000000 13.1862451 120.000000 .290231261 127.500000 .0478872294125 3102 95 135.000000 .041695504 142.500000 .040903630 150.000000 .0422966434125 3102 96 160.000000 .047405648 170.000000 .065198158 180.000000 .2623448834125 3102 97 190.000000 31.7843412 200.000000 .106103911 210.000000 .0538213784125 3102 98 220.000000 .045463565 230.000000 .053186639 240.000000 1.265151084125 3102 99 255.000000 .040834466 270.000000 .039300932 280.000000 .0402928744125 3102 100 300.000000 .616364223 320.000000 4.51525875 340.000000 .0842124844125 3102 101 360.000000 18.5996508 380.000000 1.22493655 400.000000 .0546833404125 3102 102 425.000000 .044291672 450.000000 1.39397673 475.000000 .1741005274125 3102 103 500.000000 .672458442 525.000000 .039233302 550.000000 .0392581204125 3102 104 575.000000 .184176069 600.000000 .597810156 630.000000 .5944352414125 3102 105 660.000000 1.13908614 690.000000 .116156873 720.000000 6.153567244125 3102 106 760.000000 .055145941 800.000000 .046988458 840.000000 .0513350804125 3102 107 880.000000 .294282959 920.000000 6.98986040 960.000000 .3926341014125 3102 108 1000.00000 13.4970585 1050.00000 .085749896 1100.00000 3.961236184125 3102 109 1150.00000 5.46338351 1200.00000 1.03433461 1275.00000 .6031906444125 3102 110 1350.00000 2.17525549 1425.00000 3.62240425 1500.00000 1.954560134125 3102 111 1600.00000 .920363308 1700.00000 .644013507 1800.00000 3.544938904125 3102 112 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0.000000+0 0.000000+0 1 5 210 20412540 16 4 1.000000-5 8.900000+6 9.250000+6 9.500000+6 9.750000+6 1.000000+7412540 16 5 1.050000+7 1.100000+7 1.150000+7 1.200000+7 1.250000+7 1.340000+7412540 16 6 1.400000+7 1.450000+7 1.500000+7 1.600000+7 1.700000+7 1.800000+7412540 16 7 1.900000+7 2.000000+7 1.841580-2 0.000000+0 0.000000+0 0.000000+0412540 16 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0412540 16 9 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0412540 16 10 0.000000+0 0.000000+0 0.000000+0 1.841580-2-5.728770-4 1.657170-3412540 16 11 1.455050-3 7.578240-4 1.487120-4-1.526500-4-2.597930-4-2.584590-4412540 16 12 -1.688160-4-4.368440-5 4.388290-5 1.098510-4 1.738450-4 1.938990-4412540 16 13 1.316680-4-1.410550-5-2.449450-4 6.395990-4 2.205000-4 1.633780-4412540 16 14 1.415700-4 1.287930-4 1.149910-4 9.899660-5 8.228950-5 6.007320-5412540 16 15 3.972790-5 2.775870-5 1.927900-5 1.116450-5 7.840190-6 1.330400-5412540 16 16 2.737810-5 5.011810-5 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1.360900-5 9.229420-6 1.558390-5 3.223920-5 5.907260-5 8.921230-5412540 16 30 6.886480-5 4.947110-5 3.769610-5 2.916110-5 2.062300-5 1.620220-5412540 16 31 1.989420-5 3.142810-5 5.073680-5 5.863580-5 4.779660-5 4.103260-5412540 16 32 3.592850-5 3.043560-5 2.674340-5 2.736810-5 3.221500-5 4.126730-5412540 16 33 4.558070-5 4.353010-5 4.191100-5 3.978220-5 3.756860-5 3.619940-5412540 16 34 3.571340-5 3.612990-5 4.487720-5 4.532580-5 4.556720-5 4.487230-5412540 16 35 4.310650-5 4.037530-5 3.671940-5 4.788290-5 4.994220-5 5.089690-5412540 16 36 4.958410-5 4.615120-5 4.065950-5 5.501730-5 5.869390-5 5.956310-5412540 16 37 5.761910-5 5.296130-5 6.734270-5 7.336690-5 7.759760-5 7.999730-5412540 16 38 8.807950-5 1.031920-4 1.196930-4 1.348650-4 1.722210-4 2.385330-4412540 16 39 412540 099999 4125 0 0 0 0 0 0 0 4.51030E+4 1.02021E+2 0 0 34 104525 1451 1 0.0 0.0 0 0 0 64525 1451 2 1.00000E+0 2.00000E+7 0 0 10 20024525 1451 3 3.00000E+2 0.0 1 0 179 54525 1451 4 45-Rh-103 FEI EVAL-Mar03 K.I.Zolotarev 4525 1451 5 DIST-Feb2004 4525 1451 6 ----IRDF-2002 MATERIAL 4525 4525 1451 7 -----INCIDENT NEUTRON DATA 4525 1451 8 ------ENDF-6 FORMAT 4525 1451 9 ***************************************************************** 4525 1451 10 45-Rh-103 FEI EVAL-Mar03 K.I.Zolotarev 4525 1451 11 DIST-Apr03 4525 1451 12 ----BROND-2 MATERIAL 4525 4525 1451 13 -----INCIDENT NEUTRON DATA 4525 1451 14 ------ENDF-6 FORMAT 4525 1451 15 ------Russian Reactor Dosimetry File RRDF-2002 4525 1451 16 ***************************************************************** 4525 1451 17 ----- MF=3 MT=51 ----- 4525 1451 18 For the IRDF-2002 file this reaction was converted at IAEA/NDS 4525 1451 19 The reaction MF/MT=3/51 was converted to MF/MT=10/ 4 4525 1451 20 The corresponding co-variance files were also converted 4525 1451 21 The reaction MF/MT=33/51 was converted to MF/MT=40/ 4 4525 1451 22 ***************************************************************** 4525 1451 23 Authors of evaluation: K.I.Zolotarev and P.K.Zolotarev 4525 1451 24 ***************************************************************** 4525 1451 25 MF=3 4525 1451 26 MT= 51 - Rh103(n,n')Rh103m reaction 4525 1451 27 ------------------------------------- 4525 1451 28 In the section MT= 51 are given evaluated cross section data 4525 1451 29 of neutron inelastic excitation of the 56.12-min isomeric state 4525 1451 30 in the rhodium-103 in the energy range from threshold to 20 MeV. 4525 1451 31 The isomeric level has energy 39.756 keV with spin and parity 4525 1451 32 7/2+ [1]. 4525 1451 33 Microscopic experimental data [4-13] were analyzed in the 4525 1451 34 process of preparation of input data base for the evaluation of 4525 1451 35 cross sections and their uncertainty for the Rh-103(n,n')Rh-103m 4525 1451 36 reaction. During this procedure all experimental data if it was 4525 1451 37 possible were corrected to the new recommended cross section data 4525 1451 38 for monitor reactions used in the measurements and to the new re- 4525 1451 39 commended decay data. Decay data for the rhodium-103m were taken 4525 1451 40 from ref. [2]. Recommended cross sections for the reaction 4525 1451 41 In-115(n,n')In-115m used as a monitor in the measurements [4], 4525 1451 42 [5] and [7] were taken from ref. [3]. 4525 1451 43 Special correction was done with experimental data [4], [6]. 4525 1451 44 Data of Cross [4] for the neutron energies 2.20 - 2.86 MeV 4525 1451 45 were renormalized to the measured by Paulsen et al. cross section 4525 1451 46 value (994.0+-45.7) mb at 2.60 MeV [8]. 4525 1451 47 Data of Santry and Butler [6] obtained in the measurements 4525 1451 48 with D(d,n)He3 source in the energy range 5.00-13.58 MeV were re- 4525 1451 49 normalized to the integral of Miach et al. experimental data [10] 4525 1451 50 in the energy interval 6.0-12.0 MeV, Fc= 0.90190 . Cross sections 4525 1451 51 measured by Santry and Butler with using Li7(p,n)Be7, T(p,n)He3 4525 1451 52 and T(d,n)He4 neutron sources were not corrected. 4525 1451 53 Excitation function for the Rh-103(n,n')Rh-103m reaction in 4525 1451 54 the energy region from threshold to 20 MeV was evaluated by means 4525 1451 55 of statistical analysis of experimental cross section data [4-10].4525 1451 56 Experimental cross section data [11-13] were rejected due to 4525 1451 57 their discrepancy with the main bulk of experimental data [4-10]. 4525 1451 58 In the rejected experiment [11] the cross section value was deter-4525 1451 59 mined only in a one energy point 14.20 MeV. 4525 1451 60 Statistical analysis of input cross section data was carried 4525 1451 61 out by means of PADE-2 code [14]. Rational function was used as 4525 1451 62 the model function [15]. 4525 1451 63 U-235 thermal fission [16] and Cf-252 spontaneous fission 4525 1451 64 neutron spectra [17] averaged cross sections calculated from the 4525 1451 65 the evaluated Rh-103(n,n')Rh-103m excitation function are the 4525 1451 66 following: 4525 1451 67 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4525 1451 68 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 4525 1451 69 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4525 1451 70 U-235 neutron fission ³ 715.85 ³ 702.16+-28.08 [18] 4525 1451 71 ³ ³ 721.16+-38.65 [19] 4525 1451 72 ³ ³ 670.73+-52.05 [20] 4525 1451 73 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4525 1451 74 Cf-252 spont. fission ³ 724.83 ³ 620.82+-67.17 [21] 4525 1451 75 ³ ³ 813.29+-24.24 [22] 4525 1451 76 ³ ³ 813.22+-24.15 [22] 4525 1451 77 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4525 1451 78 Integral experimental data [18-22] were corrected to the new 4525 1451 79 recommended cross sections for monitor reactions from ref. [23] 4525 1451 80 and [24]. 4525 1451 81 4525 1451 82 MT=33 4525 1451 83 MT= 51 -(n,n') cross section cov. matrix 4525 1451 84 ---------------------------------------- 4525 1451 85 Uncertainties in the evaluated excitation function for the 4525 1451 86 reaction Rh-103(n,n')Rh-103m are given in the form of relative 4525 1451 87 covariance matrix for the 35-neutron energy groups (LB=5). Cova- 4525 1451 88 riance matrix of uncertainties was calculated simultaneously with 4525 1451 89 recommended cross section data by means of PADE-2 code. 4525 1451 90 Eigenvalues of the 6-th digits relative covariance matrix 4525 1451 91 given in the 33-file are the following: 4525 1451 92 4525 1451 93 4.63337E-06 4.70345E-06 4.81607E-06 4.97060E-06 4525 1451 94 5.13303E-06 5.31401E-06 5.53768E-06 5.79915E-06 4525 1451 95 6.08110E-06 6.37403E-06 6.83267E-06 7.59079E-06 4525 1451 96 8.42883E-06 9.54383E-06 1.23066E-05 1.44930E-05 4525 1451 97 2.21073E-05 4.79767E-05 9.57950E-05 1.42110E-04 4525 1451 98 1.77925E-04 2.06906E-04 5.62715E-04 1.57661E-03 4525 1451 99 1.64672E-03 1.73445E-03 2.10901E-03 2.41391E-03 4525 1451 100 2.81682E-03 8.17990E-03 1.45397E-02 1.82324E-02 4525 1451 101 5.39402E-02 1.07397E-01 1.22586E-01 4525 1451 102 4525 1451 103 References: 4525 1451 104 1. R.B.Firestone Table of Isotopes, Eighth edition, Vol. 1, 4525 1451 105 John Wiley & Sons, Inc., New York, 1995 4525 1451 106 2. H.Vonach et al. Proc. Inter. Conf. on Nuclear Data for 4525 1451 107 Science and Technology Gatlinburg, Tennessee, USA May 9-13, 4525 1451 108 1994, v.1, pp.278-280 ; 4525 1451 109 U.Schotzig PTB, Braunschweig, FRG, 1994 4525 1451 110 3. K.I.Zolotarev, F.K.Zolotarev RRDF-2002, MAT=4921, IPPE, 4525 1451 111 Obninsk, evaluated March 2003 4525 1451 112 4. W.G.Cross Physics in Canada, v.19, p.39, June 1963 4525 1451 113 5. A.Pazsit, J.Csikai Sov. Journal of Nuclear Physics, v.15, 4525 1451 114 p.232, 1972. 4525 1451 115 6. D.C.Santry, J.P.Butler Canadian Journal of Physics, v.52, 4525 1451 116 p.1421, August 1974 4525 1451 117 7. A.Pazsit, G.Peto, J.Csikai, I.Jozsa, J.Bacso 4525 1451 118 J. of Applied Radiation and Isotopes, v.26, p.621, Oct. 1975 4525 1451 119 8. A.Paulsen et al. Nucl. Sci. Eng., v.76, p.331, 1980 4525 1451 120 9. Li Jianwei et al. Proc. of an Int. Conf. on Nuclear Data for 4525 1451 121 Science and Technology, 30 May - 3 June 1988, Mito, Japan, 4525 1451 122 Saikon Publishing Co., LTD, p.315, 1988 4525 1451 123 10. M.M.H.Miah, B.Strohmaier, H.Vonach, W.Mannhart, D.Schmidt 4525 1451 124 Phys. Rev., pt. C, v.54, No.1, p.222, July 1996 4525 1451 125 11. W.Nagel J. Nucl. Energ., v.20, p.475, June 1966 4525 1451 126 12. I.Kimura et al. Journal of Nucl. Science and Technology, 4525 1451 127 v.6, p.485, September 1969 4525 1451 128 13. E.Barnard, D.Reitmann Nucl. Phys., v.A303, p.27, June 1978 4525 1451 129 14. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 4525 1451 130 15. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 4525 1451 131 st's Meeting on Evaluation and Processing of Covariance Data, 4525 1451 132 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 4525 1451 133 16. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 4525 1451 134 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 4525 1451 135 17. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 4525 1451 136 18. E.I.Grigor'ev et al. 6th All Union Conference on Neutron 4525 1451 137 Physics, Kiev, 2-6 October 1983, v.3, p.187, Moscow 1984 4525 1451 138 19. O.Horibe et al. Proc. of Conf.: 50 Years with Nuclear Fission,4525 1451 139 Washington D.C., 25-28 April 1989, v.2, p.923 ; 4525 1451 140 O.Horibe, H.Chatani Proc. of Int. Conf. Nuclear Data for Sci.4525 1451 141 and Technology, Julich, FRG, 13-17 May 1991, Springer-Verlag, 4525 1451 142 Berlin Heidelberg, 1992, p.68 4525 1451 143 20. K.Kobayashi, T.Kobayashi Progress Report NEANDC(J)-155/U, 4525 1451 144 p.52, August 1990 4525 1451 145 21. G.J.Kirouac et al. Report - 4005, Knolls Atomic Power Lab., 4525 1451 146 January 1974 4525 1451 147 22. G.P.Lamaze et al. Nucl. Sci. Eng., v.100, p.43, Sept. 1988 4525 1451 148 23. W.Mannhart Prog. Report INDC(Ger)-045, pp.40-43, June 1999 4525 1451 149 24. W.Mannhart Validation of Differential Cross Sections with 4525 1451 150 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 4525 1451 151 September 2002 4525 1451 152 ***************************************************************** 4525 1451 153 File 2 added to the pointwise file containing only the effective 4525 1451 154 scattering radius with no resonance parameters given. 4525 1451 155 Taken from ENDF/B-VI 4525 1451 156 4525 1451 157 ***************************************************************** 4525 1451 158 ***************** Program LINEAR (VERSION 2002-1) ***************4525 1451 159 For All Data Greater than 1.0000E-10 barns in Absolute Value 4525 1451 160 Data Linearized to Within an Accuracy of .100000000 per-cent 4525 1451 161 ***************** Program SIGMA1 (VERSION 2002-1) ***************4525 1451 162 Data Doppler Broadened to 300.000000 Kelvin 4525 1451 163 for All Data Greater than 1.0000E-10 barns in Absolute Value 4525 1451 164 Data Linearized to Within an Accuracy pf .100000000 per-cent 4525 1451 165 ***************** Program FIXUP (Version 2002-1) ****************4525 1451 166 Corrected ZA/AWR in All Sections-----------------------------Yes 4525 1451 167 Corrected Thresholds-----------------------------------------Yes 4525 1451 168 Extended Cross Sections to 20 MeV----------------------------No 4525 1451 169 Allow Cross Section Deletion---------------------------------No 4525 1451 170 Allow Cross Section Reconstruction---------------------------No 4525 1451 171 Make All Cross Sections Non-Negative-------------------------Yes 4525 1451 172 Delete Energies Not in Ascending Order-----------------------Yes 4525 1451 173 Deleted Duplicate Points-------------------------------------Yes 4525 1451 174 Check for Ascending MAT/MF/MT Order--------------------------Yes 4525 1451 175 Check for Legal MF/MT Numbers--------------------------------Yes 4525 1451 176 Allow Creation of Missing Sections---------------------------No 4525 1451 177 Allow Insertion of Energy Points-----------------------------No 4525 1451 178 Create Uniform Energy Grid-----------------------------------No 4525 1451 179 Delete Section if Cross Section =0 at All Energies-----------Yes 4525 1451 180 ***************** Program GROUPIE (VERSION 2002-1) **************4525 1451 181 Unshielded Group Averages Using 640 Groups 4525 1451 182 Weighting Spectrum: Flat (Constant) Spectrum 4525 1451 183 1 451 188 14525 1451 184 2 151 4 14525 1451 185 8 4 2 14525 1451 186 10 4 88 14525 1451 187 40 4 122 14525 1451 188 4525 1 099999 4525 0 0 0 4.51030E+4 1.02021E+2 0 0 1 04525 2151 1 4.510300+4 1.000000+0 0 0 1 04525 2151 2 1.000000-5 4.115900+3 0 0 0 04525 2151 3 5.000000-1 6.560000-1 0 0 0 04525 2151 4 4525 2 099999 4525 0 0 0 4.51030E+4 1.02021E+2 0 0 1 14525 8 4 1 4.51030E+4 0.000000+0 10 0 0 04525 8 4 2 4525 8 099999 4525 0 0 0 4.51030E+4 1.02021E+2 0 0 1 0452510 4 1 0.0 -3.97560E+4 45103 0 1 255452510 4 2 255 1 452510 4 3 40000.0000 .000131876 42500.0000 .000427871 45000.0000 .000724345452510 4 4 47500.0000 .001020819 50000.0000 .001317294 52500.0000 .001613768452510 4 5 55000.0000 .001910242 57500.0000 .002206716 60000.0000 .002532837452510 4 6 63000.0000 .002888606 66000.0000 .003244375 69000.0000 .003600144452510 4 7 72000.0000 .004015207 76000.0000 .004489566 80000.0000 .004963924452510 4 8 84000.0000 .005438283 88000.0000 .005912641 92000.0000 .006387000452510 4 9 96000.0000 .006861358 100000.000 .007395012 105000.000 .007987960452510 4 10 110000.000 .008580908 115000.000 .009173856 120000.000 .011379589452510 4 11 127500.000 .015198106 135000.000 .019016624 142500.000 .022835141452510 4 12 150000.000 .027155800 160000.000 .031978600 170000.000 .036801400452510 4 13 180000.000 .041624200 190000.000 .046447000 200000.000 .051090690452510 4 14 210000.000 .055555270 220000.000 .060019850 230000.000 .064484430452510 4 15 240000.000 .070033483 255000.000 .076286950 270000.000 .081392600452510 4 16 280000.000 .087519380 300000.000 .095305320 320000.000 .102708160452510 4 17 340000.000 .110028295 360000.000 .116852200 380000.000 .123593400452510 4 18 400000.000 .131003000 425000.000 .139081000 450000.000 .147617250452510 4 19 475000.000 .156611750 500000.000 .167446250 525000.000 .180120750452510 4 20 550000.000 .196934000 575000.000 .217886000 600000.000 .248218100452510 4 21 630000.000 .288511267 660000.000 .336357000 690000.000 .385875133452510 4 22 720000.000 .440636700 760000.000 .491837800 800000.000 .530516200452510 4 23 840000.000 .558433425 880000.000 .578712300 920000.000 .594550300452510 4 24 960000.000 .607908400 1000000.00 .628120500 1100000.00 .654950833452510 4 25 1200000.00 .680806500 1300000.00 .706662167 1400000.00 .731736500452510 4 26 1500000.00 .756029500 1600000.00 .779401000 1700000.00 .801851000452510 4 27 1800000.00 .823285500 1900000.00 .843704500 2000000.00 .863116000452510 4 28 2100000.00 .881520000 2200000.00 .898753833 2300000.00 .914817500452510 4 29 2400000.00 .930881167 2500000.00 .945705167 2600000.00 .959289500452510 4 30 2700000.00 .972873833 2800000.00 .985413333 2900000.00 .996908000452510 4 31 3000000.00 1.00840267 3100000.00 1.01891625 3200000.00 1.02844875452510 4 32 3300000.00 1.03798125 3400000.00 1.04751375 3500000.00 1.05618625452510 4 33 3600000.00 1.06399875 3700000.00 1.07181125 3800000.00 1.07962375452510 4 34 3900000.00 1.08670417 4000000.00 1.09305250 4100000.00 1.09940083452510 4 35 4200000.00 1.10574917 4300000.00 1.11209750 4400000.00 1.11844583452510 4 36 4500000.00 1.12419500 4600000.00 1.12934500 4700000.00 1.13449500452510 4 37 4800000.00 1.13964500 4900000.00 1.14479500 5000000.00 1.14994500452510 4 38 5100000.00 1.15509500 5200000.00 1.16024500 5300000.00 1.16539500452510 4 39 5400000.00 1.17054500 5500000.00 1.17544306 5600000.00 1.18008917452510 4 40 5700000.00 1.18473528 5800000.00 1.18938139 5900000.00 1.19402750452510 4 41 6000000.00 1.19867361 6100000.00 1.20331972 6200000.00 1.20796583452510 4 42 6300000.00 1.21261194 6400000.00 1.21725806 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14300000.0 .281220500452510 4 69 14400000.0 .275846750 14500000.0 .270816250 14600000.0 .266083250452510 4 70 14700000.0 .261647750 14800000.0 .257527167 14900000.0 .253721500452510 4 71 15000000.0 .249915833 15100000.0 .246422000 15200000.0 .243240000452510 4 72 15300000.0 .240058000 15400000.0 .237126500 15500000.0 .234445500452510 4 73 15600000.0 .231764500 15700000.0 .229284167 15800000.0 .227004500452510 4 74 15900000.0 .224724833 16000000.0 .222629125 16100000.0 .220717375452510 4 75 16200000.0 .218805625 16300000.0 .216893875 16400000.0 .215159000452510 4 76 16500000.0 .213601000 16600000.0 .212043000 16700000.0 .210485000452510 4 77 16800000.0 .208927000 16900000.0 .207520000 17000000.0 .206264000452510 4 78 17100000.0 .205008000 17200000.0 .203752000 17300000.0 .202496000452510 4 79 17400000.0 .201240000 17500000.0 .200100429 17600000.0 .199077286452510 4 80 17700000.0 .198054143 17800000.0 .197031000 17900000.0 .196007857452510 4 81 18000000.0 .194984714 18100000.0 .193961571 18200000.0 .193024167452510 4 82 18300000.0 .192172500 18400000.0 .191320833 18500000.0 .190469167452510 4 83 18600000.0 .189617500 18700000.0 .188765833 18800000.0 .187914167452510 4 84 18900000.0 .187062500 19000000.0 .186210833 19100000.0 .185413667452510 4 85 19200000.0 .184671000 19300000.0 .183928333 19400000.0 .183185667452510 4 86 19500000.0 .182443000 19600000.0 .181700333 19700000.0 .180957667452510 4 87 19800000.0 .180215000 19900000.0 .179472333 20000000.0 0.0 452510 4 88 452510 099999 4525 0 0 0 4.51030E+4 1.02021E+2 0 0 1 0452540 4 1 0.000000+0-3.975600+4 0 0 0 1452540 4 2 1.000000+1 0.000000+0 0 4 0 1452540 4 3 0.000000+0 0.000000+0 1 5 703 37452540 4 4 1.000000-5 4.000000+4 2.000000+5 4.000000+5 6.000000+5 8.000000+5452540 4 5 1.000000+6 1.500000+6 2.000000+6 2.500000+6 3.000000+6 3.500000+6452540 4 6 4.000000+6 4.500000+6 5.000000+6 5.500000+6 6.000000+6 6.500000+6452540 4 7 7.000000+6 7.500000+6 8.000000+6 8.500000+6 9.000000+6 9.500000+6452540 4 8 1.000000+7 1.050000+7 1.100000+7 1.150000+7 1.200000+7 1.300000+7452540 4 9 1.400000+7 1.500000+7 1.600000+7 1.700000+7 1.800000+7 1.900000+7452540 4 10 2.000000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452540 4 11 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452540 4 12 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452540 4 13 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452540 4 14 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452540 4 15 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0452540 4 16 0.000000+0 1.201660-1 3.626120-3 5.830250-3 2.406510-3 3.829450-3452540 4 17 9.421320-4 1.744590-3 2.795830-3 3.312800-3 3.411860-3 3.250220-3452540 4 18 2.951990-3 2.607590-3 2.280380-3 2.013300-3 1.832210-3 1.747890-3452540 4 19 1.754810-3 1.828740-3 1.924420-3 1.979780-3 1.934180-3 1.764500-3452540 4 20 1.522800-3 1.330280-3 1.305230-3 1.474590-3 1.853920-3 1.977210-3452540 4 21 1.349570-3 4.168760-4-3.002750-4-5.661320-4-3.496380-4 3.015310-4452540 4 22 4.977370-3 2.720970-3 2.068350-3 1.881520-3 1.983850-3 1.918150-3452540 4 23 1.869060-3 1.891870-3 1.958280-3 2.036100-3 2.099860-3 2.130840-3452540 4 24 2.116050-3 2.047420-3 1.922130-3 1.743840-3 1.525610-3 1.293820-3452540 4 25 1.091910-3 9.788930-4 1.014690-3 1.226060-3 1.564980-3 1.899910-3452540 4 26 2.072300-3 1.986420-3 1.447110-3 5.347710-4 3.983250-6-8.023330-5452540 4 27 2.977010-5 1.266280-4 1.141860-4-3.476890-5 3.978040-3 1.824830-3452540 4 28 1.835480-3 1.643210-3 1.509900-3 1.548560-3 1.631550-3 1.715330-3452540 4 29 1.779860-3 1.814290-3 1.812430-3 1.770950-3 1.688910-3 1.568090-3452540 4 30 1.414200-3 1.238960-3 1.062830-3 9.173610-4 8.435190-4 8.807400-4452540 4 31 1.042860-3 1.290290-3 1.526610-3 1.640340-3 1.567330-3 1.166990-3452540 4 32 4.944710-4 8.494070-5-1.223270-5 3.449800-5 8.571080-5 7.622970-5452540 4 33 -1.302230-5 3.335450-3 1.865850-3 1.320780-3 1.334180-3 1.390290-3452540 4 34 1.451910-3 1.509870-3 1.554570-3 1.578240-3 1.575150-3 1.541460-3452540 4 35 1.475190-3 1.376710-3 1.249730-3 1.103140-3 9.534420-4 8.269430-4452540 4 36 7.587760-4 7.840130-4 9.172220-4 1.127730-3 1.334760-3 1.441390-3452540 4 37 1.388440-3 1.050250-3 4.598470-4 8.317440-5-1.962000-5 1.285200-5452540 4 38 5.886810-5 5.874010-5-6.850400-6 2.935500-3 1.948970-3 1.404000-3452540 4 39 1.306000-3 1.344960-3 1.428310-3 1.513750-3 1.577960-3 1.607040-3452540 4 40 1.592890-3 1.531610-3 1.423490-3 1.273950-3 1.095770-3 9.119690-4452540 4 41 7.582380-4 6.808130-4 7.237980-4 9.020240-4 1.170360-3 1.421600-3452540 4 42 1.534730-3 1.444670-3 1.011040-3 3.314790-4-2.048830-5-3.774030-5452540 4 43 6.704390-5 1.361820-4 1.034210-4-4.532930-5 2.185350-3 1.745000-3452540 4 44 1.484750-3 1.385400-3 1.380060-3 1.419750-3 1.470620-3 1.509200-3452540 4 45 1.519370-3 1.490590-3 1.417430-3 1.300150-3 1.146670-3 9.756850-4452540 4 46 8.197340-4 7.246800-4 7.391190-4 8.883700-4 1.141090-3 1.399790-3452540 4 47 1.541910-3 1.489490-3 1.091420-3 4.000170-4-6.305970-6-6.918640-5452540 4 48 1.719700-5 8.992070-5 7.228520-5-5.799610-5 1.705580-3 1.595370-3452540 4 49 1.530820-3 1.492930-3 1.468820-3 1.447610-3 1.420890-3 1.382350-3452540 4 50 1.327560-3 1.254000-3 1.161690-3 1.054450-3 9.418340-4 8.412030-4452540 4 51 7.780520-4 7.807980-4 8.663420-4 1.019870-3 1.186120-3 1.289640-3452540 4 52 1.274940-3 1.030670-3 5.329380-4 1.499900-4-1.318120-5-3.406270-5452540 4 53 -5.586790-6 1.728610-5 1.067630-5 1.634360-3 1.625160-3 1.607390-3452540 4 54 1.574910-3 1.528640-3 1.468740-3 1.395220-3 1.308330-3 1.209170-3452540 4 55 1.100440-3 9.876960-4 8.808630-4 7.955860-4 7.526840-4 7.727340-4452540 4 56 8.635330-4 1.004710-3 1.144140-3 1.218310-3 1.185790-3 9.517920-4452540 4 57 5.073870-4 1.702880-4 1.966640-5-9.743300-6 7.648470-6 2.860580-5452540 4 58 3.445390-5 1.686590-3 1.698690-3 1.689110-3 1.652020-3 1.589090-3452540 4 59 1.501910-3 1.392580-3 1.264400-3 1.123050-3 9.780430-4 8.445670-4452540 4 60 7.445890-4 7.049900-4 7.491080-4 8.799410-4 1.062630-3 1.225610-3452540 4 61 1.292150-3 1.222690-3 9.173280-4 4.205910-4 1.117470-4 2.198870-5452540 4 62 3.810570-5 6.754520-5 6.709010-5 2.377900-5 1.767300-3 1.787600-3452540 4 63 1.778600-3 1.732110-3 1.648380-3 1.529190-3 1.378700-3 1.204750-3452540 4 64 1.020930-3 8.488400-4 7.195540-4 6.705770-4 7.337420-4 9.118860-4452540 4 65 1.155540-3 1.366660-3 1.444020-3 1.340990-3 9.399110-4 3.401500-4452540 4 66 3.115010-5 3.701460-6 7.870690-5 1.295250-4 1.052400-4-3.667290-6452540 4 67 1.861870-3 1.882680-3 1.864080-3 1.796830-3 1.681110-3 1.520660-3452540 4 68 1.324290-3 1.108380-3 8.998800-4 7.381710-4 6.716550-4 7.426720-4452540 4 69 9.579980-4 1.258590-3 1.523090-3 1.624410-3 1.503210-3 1.018640-3452540 4 70 3.048970-4-3.417320-5-2.491860-5 9.680240-5 1.687500-4 1.281170-4452540 4 71 -3.466650-5 1.953700-3 1.962510-3 1.921520-3 1.822400-3 1.667480-3452540 4 72 1.465140-3 1.232610-3 9.995790-4 8.107960-4 7.228870-4 7.879580-4452540 4 73 1.019900-3 1.357890-3 1.666730-3 1.798970-3 1.680060-3 1.145900-3452540 4 74 3.346320-4-6.084060-5-5.413060-5 8.742170-5 1.740050-4 1.296000-4452540 4 75 -5.921130-5 2.018400-3 2.004860-3 1.934740-3 1.802360-3 1.614240-3452540 4 76 1.386200-3 1.147330-3 9.433540-4 8.345250-4 8.798450-4 1.102280-3452540 4 77 1.448040-3 1.781360-3 1.944550-3 1.848350-3 1.309570-3 4.359200-4452540 4 78 -3.516650-5-7.449720-5 5.289760-5 1.437700-4 1.099710-4-6.921970-5452540 4 79 2.040740-3 2.005170-3 1.912560-3 1.760710-3 1.561740-3 1.340470-3452540 4 80 1.138540-3 1.013920-3 1.027960-3 1.213110-3 1.531140-3 1.860530-3452540 4 81 2.046820-3 1.989450-3 1.493570-3 6.047000-4 4.891560-5-7.673680-5452540 4 82 1.371240-6 8.420380-5 7.466270-5-5.785160-5 2.030060-3 1.988070-3452540 4 83 1.894760-3 1.751080-3 1.574090-3 1.396040-3 1.266060-3 1.241910-3452540 4 84 1.364030-3 1.616560-3 1.907010-3 2.099450-3 2.088640-3 1.678040-3452540 4 85 8.271910-4 1.904990-4-5.195030-5-5.417730-5 7.952040-6 3.346140-5452540 4 86 -1.934510-5 2.026430-3 2.007020-3 1.944500-3 1.840030-3 1.711650-3452540 4 87 1.592980-3 1.531150-3 1.570710-3 1.721980-3 1.934050-3 2.105380-3452540 4 88 2.136050-3 1.839400-3 1.079100-3 3.806240-4 7.422110-6-9.667630-5452540 4 89 -6.646090-5-7.827400-8 5.032560-5 2.093200-3 2.129750-3 2.125450-3452540 4 90 2.075490-3 1.991990-3 1.903760-3 1.852790-3 1.874580-3 1.967840-3452540 4 91 2.079970-3 2.129570-3 1.951850-3 1.324730-3 6.001410-4 1.053960-4452540 4 92 -1.067670-4-1.161000-4-1.004330-5 1.517680-4 2.296170-3 2.410830-3452540 4 93 2.468920-3 2.452110-3 2.363350-3 2.232360-3 2.112150-3 2.051350-3452540 4 94 2.056640-3 2.080990-3 1.991540-3 1.516680-3 8.163850-4 2.383610-4452540 4 95 -6.646880-5-1.167200-4 1.891140-5 2.788530-4 2.672770-3 2.860640-3452540 4 96 2.948500-3 2.902960-3 2.729070-3 2.481360-3 2.246960-3 2.093980-3452540 4 97 2.025970-3 1.946320-3 1.600080-3 9.808020-4 3.887410-4 3.325860-5452540 4 98 -4.995920-5 9.539570-5 4.136770-4 3.207960-3 3.435520-3 3.495000-3452540 4 99 3.349050-3 3.028830-3 2.633510-3 2.280710-3 2.036640-3 1.834010-3452540 4 100 1.526700-3 1.031610-3 5.177390-4 1.802330-4 8.386130-5 2.111010-4452540 4 101 5.209780-4 3.846890-3 4.078820-3 4.065070-3 3.778340-3 3.287360-3452540 4 102 2.728170-3 2.229490-3 1.728800-3 1.286970-3 9.106850-4 5.629970-4452540 4 103 3.263650-4 2.481130-4 3.284120-4 5.472190-4 4.558340-3 4.791720-3452540 4 104 4.690670-3 4.237160-3 3.533920-3 2.750150-3 1.782710-3 9.594800-4452540 4 105 6.041830-4 4.544830-4 3.836850-4 3.591780-4 3.767060-4 4.355290-4452540 4 106 5.379550-3 5.621120-3 5.396270-3 4.706480-3 3.711170-3 2.207740-3452540 4 107 7.515130-4 2.027880-4 1.638560-4 2.563050-4 3.091530-4 2.787680-4452540 4 108 1.688620-4 6.326250-3 6.518650-3 6.081040-3 5.085380-3 3.181080-3452540 4 109 9.695790-4-6.047840-5-2.210550-4-7.551840-5 4.549650-5 2.181970-5452540 4 110 -1.665760-4 7.262800-3 7.312700-3 6.627790-3 4.695680-3 1.873840-3452540 4 111 1.345200-4-4.562920-4-4.594530-4-3.217850-4-2.598870-4-3.491170-4452540 4 112 8.021830-3 7.929730-3 6.477870-3 3.489070-3 1.036030-3-2.102110-4452540 4 113 -5.656340-4-4.887160-4-2.780690-4-8.214510-5 8.624160-3 8.079560-3452540 4 114 5.538600-3 2.666940-3 7.744250-4-2.478930-5-7.501780-5 2.929000-4452540 4 115 8.666250-4 9.222250-3 8.299370-3 5.838530-3 3.614610-3 2.386360-3452540 4 116 2.132820-3 2.604510-3 3.570530-3 1.032610-2 9.947210-3 8.723800-3452540 4 117 7.726450-3 7.373060-3 7.680650-3 8.542170-3 1.246280-2 1.317360-2452540 4 118 1.318840-2 1.302470-2 1.298030-2 1.315830-2 1.597610-2 1.695770-2452540 4 119 1.713040-2 1.683090-2 1.633520-2 1.899860-2 1.933620-2 1.899850-2452540 4 120 1.818580-2 2.027880-2 1.992020-2 1.919880-2 2.027840-2 1.978060-2452540 4 121 2.040580-2 452540 4 122 452540 099999 4525 0 0 0 0 0 0 0 4.71090E+4 1.07969E+2 0 0 34 104731 1451 1 0.0 0.0 0 0 0 64731 1451 2 1.00000E+0 2.00000E+7 0 0 10 20024731 1451 3 3.00000E+2 0.0 1 0 138 54731 1451 4 47-Ag-109 CNDC EVAL-JUN91 Z.X.ZHAO 4731 1451 5 DIST-FEB2004 4731 1451 6 ----IRDF-2002 MATERIAL 4731 4731 1451 7 -----INCIDENT NEUTRON DATA 4731 1451 8 ------ENDF-6 FORMAT 4731 1451 9 ================================================================ 4731 1451 10 47-AG-109 CNDC EVAL-JUN91 Z.X.ZHAO 4731 1451 11 DIST- 4731 1451 12 ----ENDF/A-6 MATERIAL 4731 4731 1451 13 -----INCIDENT NEUTRON DATA 4731 1451 14 ------ENDF-6 FORMAT 4731 1451 15 * * * * * * *4731 1451 16 THIS EVALUATION WAS PERFORMED BASED ON THE MEASURED DATA 4731 1451 17 AVAILABLE AND THE CALCULATIONS OF THE SYSTEMATICS AND THEORY. 4731 1451 18 FILE 2 RESONANCE PARAMETER 4731 1451 19 ONLY POTENTIAL SCATTERING RADIUS R' OF REF.[1] IS GIVEN. 4731 1451 20 FILE 10 PRODUCTION CROSS SECTION OF RADIOACTIVE NUCLIDE 4731 1451 21 MT=102 109-AG(N,GAMMA)110M-AG REACTION 4731 1451 22 THE PRODUCTION CROSS SECTION OF ISOMER STATE OF 110-AG 4731 1451 23 ARE OBTAINED BY 4731 1451 24 C.S.(M) = C.S.(M+G)*R 4731 1451 25 WHERE C.S.(M+G) REPRESENTS THE TOTAL CAPTURE CROSS SECTION 4731 1451 26 AND R IS THE RATIO OF THE CAPTURE CROSS SECTION TO ISOMER 4731 1451 27 STATE AND THE TOTAL CAPTURE CROSS SECTION. THE VALUES OF 4731 1451 28 R WERE CALCULATED BY USING UNIFY2[2] AND NORMALIZED TO 4731 1451 29 THE LIMITED DATA FOR C.S.(M) MEASURED AT THERMAL CROSS 4731 1451 30 SECTION [3-5] AND AROUND 25 KEV [6-7]. 4731 1451 31 C.S.(M+G) WAS RE-EVALUATED IN THE FOLLOWING WAY: 4731 1451 32 BELOW 2.5 KEV, C.S.(M+G) WAS CALCULATED FROM A SET 4731 1451 33 OF RESONANCE PARAMETERS OF JENDL-3 AND AVERAGED IN 4731 1451 34 SMALL ENERGY INTERVALS. 4731 1451 35 FROM 2.5 KEV TO 20 MEV, THE SYSTEMATICS OF ZHAO [8] 4731 1451 36 WAS USED TO FIT THE MEASURED DATA OF REFS.[9-10] IN THE 4731 1451 37 ENERGY RANGE FROM 2.5 KEV TO 2 MEV. 4731 1451 38 FILE 8 RADIOACTIVE DECAY DATA 4731 1451 39 THE RADIOACTIVE DECAY DATA FOR 110M-AG ARE TAKEN FROM 4731 1451 40 REFS.[11] AND [12]. 4731 1451 41 FILE 40 COVARIANCE DATA FOR FILE 10 4731 1451 42 COVARIANCE DATA OF FILE 10 CONSISTES OF TWO COMPONENTS. 4731 1451 43 ONE OF THEM COME FROM THE UNCERTAINTY OF TOTAL CAPTURE 4731 1451 44 CROSS SECTION. ANOTHER COME FROM THE NORMALIZATION ERROR 4731 1451 45 OF R=C.S.(M)/C.S.(M+G). 4731 1451 46 ================================================================ 4731 1451 47 =========== Processing done at IAEA/NDS for IRDF-2002 ========== 4731 1451 48 ================================================================ 4731 1451 49 The original file was calculated from a set of resonance 4731 1451 50 parameters taken from JENDL-3. 4731 1451 51 The JENDL-3.2 file for Ag109 was processed as follows to produce 4731 1451 52 a pointwise file. 4731 1451 53 ***************** Program LINEAR (VERSION 2002-1) ***************4731 1451 54 For All Data Greater than 1.0000E-10 barns in Absolute Value 4731 1451 55 Data Linearized to Within an Accuracy of .100000000 per-cent 4731 1451 56 ***************** Program RECENT (VERSION 2002-1) ***************4731 1451 57 for All Data Greater than 1.0000E-10 barns in Absolute Value 4731 1451 58 Data Linearized to within an Accuracy of .100000000 per-cent 4731 1451 59 ***************** Program SIGMA1 (VERSION 2002-1) ***************4731 1451 60 Data Doppler Broadened to 300.000000 Kelvin 4731 1451 61 for All Data Greater than 1.0000E-10 barns in Absolute Value 4731 1451 62 Data Linearized to Within an Accuracy pf 1.00000000 per-cent 4731 1451 63 ***************** Program FIXUP (Version 2002-1) ****************4731 1451 64 Corrected ZA/AWR in All Sections-----------------------------Yes 4731 1451 65 Corrected Thresholds-----------------------------------------No 4731 1451 66 Extended Cross Sections to 20 MeV----------------------------No 4731 1451 67 Allow Cross Section Deletion---------------------------------No 4731 1451 68 Allow Cross Section Reconstruction---------------------------No 4731 1451 69 Make All Cross Sections Non-Negative-------------------------Yes 4731 1451 70 Delete Energies Not in Ascending Order-----------------------Yes 4731 1451 71 Deleted Duplicate Points-------------------------------------Yes 4731 1451 72 Check for Ascending MAT/MF/MT Order--------------------------Yes 4731 1451 73 Check for Legal MF/MT Numbers--------------------------------Yes 4731 1451 74 Allow Creation of Missing Sections---------------------------No 4731 1451 75 Allow Insertion of Energy Points-----------------------------No 4731 1451 76 Create Uniform Energy Grid-----------------------------------No 4731 1451 77 Delete Section if Cross Section =0 at All Energies-----------Yes 4731 1451 78 4731 1451 79 The branching ratio of 0.0465 was applied to the Ag109 total 4731 1451 80 capture cross section data of the JENDL-3.2 file in the energy 4731 1451 81 range 1.0E-5 to 7.1E+3. The scaled data with the above energy 4731 1451 82 range was then edited into the original data file as used for 4731 1451 83 IRDF-90 4731 1451 84 ================================================================ 4731 1451 85 4731 1451 86 REFERENCES 4731 1451 87 [1] S. F. Mughabghab et al., BNL-325, 4th ed., Vol.(1981) 4731 1451 88 4731 1451 89 [2] Zhang Jinshang, " Semi-classical Theory Code UNIFY of 4731 1451 90 Multi-step Nuclear Reaction", to be published 4731 1451 91 4731 1451 92 [3] T. B. Ryves et al., J. Nucl. Energy, 25(1971)129 4731 1451 93 4731 1451 94 [4] J. K. Aaldijk et al., RCN-176,1972 4731 1451 95 4731 1451 96 [5] A. Simonits et al., JRN, 81(1984)369 4731 1451 97 4731 1451 98 [6] M. Sriramachandra et al., JP/A,5(1972)877 4731 1451 99 4731 1451 100 [7] W. Poenitz, EANDC(E0-66,6(1966) 4731 1451 101 4731 1451 102 [8] Zhao Zhixiang et al., Chinese Nucl. Phys., 11(1989)71 4731 1451 103 4731 1451 104 [9] R. L. Macklin, Nucl. Sci. and Eng., 82(1982)400 4731 1451 105 4731 1451 106 [10] M. Mizumoto et al., "Neutron Radiative Capture and Transm4731 1451 107 Measurements of 107-Ag and 109-Ag", Proc. Int. Conf. Nucl4731 1451 108 for Sci. and Tech., Antwerp, 1982, p.226(1983) 4731 1451 109 4731 1451 110 [11] J. K. Tuli, Nuclear Wallet Cards ,1990 4731 1451 111 4731 1451 112 [12] C. M. Lederer and V. S. shirley, Table of Isotopes, 4731 1451 113 Seventh Edition(1978) 4731 1451 114 4731 1451 115 *****************************************************************4731 1451 116 4731 1451 117 4731 1451 118 4731 1451 119 4731 1451 120 ***************** PROGRAM LINEAR (VERSION 96-1) ************* 4731 1451 121 FOR ALL DATA GREATER THAN 1.00000-10 BARNS IN ABSOLUTE VALUE 4731 1451 122 DATA LINEARIZED TO WITHIN AN ACCURACY OF 0.10000000 PER-CENT 4731 1451 123 ***************** Program FIXUP (Version 2002-1) ****************4731 1451 124 Corrected ZA/AWR in All Sections-----------------------------Yes 4731 1451 125 Corrected Thresholds-----------------------------------------Yes 4731 1451 126 Extended Cross Sections to 20 MeV----------------------------No 4731 1451 127 Allow Cross Section Deletion---------------------------------No 4731 1451 128 Allow Cross Section Reconstruction---------------------------No 4731 1451 129 Make All Cross Sections Non-Negative-------------------------Yes 4731 1451 130 Delete Energies Not in Ascending Order-----------------------Yes 4731 1451 131 Deleted Duplicate Points-------------------------------------Yes 4731 1451 132 Check for Ascending MAT/MF/MT Order--------------------------Yes 4731 1451 133 Check for Legal MF/MT Numbers--------------------------------Yes 4731 1451 134 Allow Creation of Missing Sections---------------------------No 4731 1451 135 Allow Insertion of Energy Points-----------------------------No 4731 1451 136 Create Uniform Energy Grid-----------------------------------No 4731 1451 137 Delete Section if Cross Section =0 at All Energies-----------Yes 4731 1451 138 *****************************************************************4731 1451 139 ***************** Program GROUPIE (VERSION 2002-1) **************4731 1451 140 Unshielded Group Averages Using 640 Groups 4731 1451 141 Weighting Spectrum: Flat (Constant) Spectrum 4731 1451 142 1 451 147 14731 1451 143 2 151 4 14731 1451 144 8 102 2 14731 1451 145 10 102 217 14731 1451 146 40 102 10 14731 1451 147 4731 1 099999 4731 0 0 0 4.71090E+4 1.07969E+2 0 0 1 04731 2151 1 4.710900+4 1.000000+0 0 0 1 04731 2151 2 1.000000-5 2.500000+3 0 0 0 04731 2151 3 5.000000-1 6.600000-1 0 0 0 04731 2151 4 4731 2 099999 4731 0 0 0 4.71090E+4 1.07969E+2 0 0 1 14731 8102 1 4.711000+4 0.0 10 2 0 04731 8102 2 4731 8 099999 4731 0 0 0 4.71090E+4 1.07969E+2 0 0 1 0473110102 1 6.80570E+6 6.68810E+6 47110 2 1 641473110102 2 641 1 473110102 3 .000100000 65.8829137 .000105000 64.3718842 .000110000 62.8608547473110102 4 .000115000 61.3498263 .000120000 59.7673282 .000127500 58.2032108473110102 5 .000135000 56.6406830 .000142500 55.0781536 .000150000 53.3719172473110102 6 .000160000 51.8716541 .000170000 50.4199466 .000180000 48.9682391473110102 7 .000190000 47.5717224 .000200000 46.5123774 .000210000 45.5102445473110102 8 .000220000 44.5081115 .000230000 43.5059794 .000240000 42.2613097473110102 9 .000255000 41.0758325 .000270000 40.2204986 .000280000 39.1940977473110102 10 .000300000 37.8255648 .000320000 36.5868623 .000340000 35.6335300473110102 11 .000360000 34.6903705 .000380000 33.7472116 .000400000 32.7187027473110102 12 .000425000 31.8382357 .000450000 31.0164125 .000475000 30.1945889473110102 13 .000500000 29.3727662 .000525000 28.6779698 .000550000 28.1107284473110102 14 .000575000 27.5434871 .000600000 26.9195219 .000630000 26.2388329473110102 15 .000660000 25.5726206 .000690000 25.0645516 .000720000 24.5224400473110102 16 .000760000 23.9028841 .000800000 23.2833283 .000840000 22.6665876473110102 17 .000880000 22.1733049 .000920000 21.7461036 .000960000 21.3189020473110102 18 .001000000 20.8383005 .001050000 20.3042989 .001100000 19.8093341473110102 19 .001150000 19.4297664 .001200000 18.9647759 .001275000 18.4067875473110102 20 .001350000 17.8509108 .001425000 17.4043156 .001500000 16.9551107473110102 21 .001600000 16.4417334 .001700000 15.9283562 .001800000 15.4574255473110102 22 .001900000 15.1005702 .002000000 14.7500021 .002100000 14.3994339473110102 23 .002200000 14.0488654 .002300000 13.7117119 .002400000 13.3958410473110102 24 .002550000 13.0333777 .002700000 12.7313251 .002800000 12.3688618473110102 25 .003000000 11.9573379 .003200000 11.6207172 .003400000 11.2841050473110102 26 .003600000 10.9474928 .003800000 10.6608505 .004000000 10.3995298473110102 27 .004250000 10.1092003 .004500000 9.81887067 .004750000 9.53324534473110102 28 .005000000 9.31366335 .005250000 9.11538011 .005500000 8.91709686473110102 29 .005750000 8.71881361 .006000000 8.50070207 .006300000 8.29350107473110102 30 .006600000 8.12938064 .006900000 7.96555528 .007200000 7.77442561473110102 31 .007600000 7.55599162 .008000000 7.35868335 .008400000 7.20541658473110102 32 .008800000 7.05352267 .009200000 6.90162877 .009600000 6.74973504473110102 33 .010000000 6.58673552 .010500000 6.44858051 .011000000 6.31718378473110102 34 .011500000 6.18578706 .012000000 6.02154121 .012750000 5.82813424473110102 35 .013500000 5.68058330 .014250000 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1451 2 1.00000E+0 2.00000E+7 0 0 10 20024800 1451 3 3.00000E+2 0.0 1 0 77 44800 1451 4 48-Cd- 0 BNL EVAL-MAY74 S.PEARLSTEIN(TRANS FROM U.K.) 4800 1451 5 DIST-Feb2004 4800 1451 6 ----IRDF-2002 MATERIAL 4800 4800 1451 7 -----INCIDENT NEUTRON DATA 4800 1451 8 ------ENDF-6 FORMAT 4800 1451 9 *****************************************************************4800 1451 10 48-CD- 0 BNL EVAL-MAY74 S.PEARLSTEIN(TRANS FROM U.K.) 4800 1451 11 DIST-JAN90 19900130 4800 1451 12 ----ENDF/B-VI MATERIAL 4800 4800 1451 13 -----INCIDENT NEUTRON DATA 4800 1451 14 ------ENDF-6 FORMAT 4800 1451 15 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 4800 1451 16 * * * * * * *4800 1451 17 UKNDL TO ENDF/B TRANSLATION FROM 4800 1451 18 UKNDL DATA FILE 70 RECEIVED JULY 1973 FROM CCDN 4800 1451 19 4800 1451 20 TRANSLATION OF UKNDL DFN-70 INTO ENDF FORMAT 4800 1451 21 4800 1451 22 RESONANCE REGION EVALUATED BY MF JAMES IN 1969 4800 1451 23 FICTITIOUS CD-111 RESONANCE AT -0.5 EV TO ADJUST CROSS SECTION 4800 1451 24 IN EV RANGE AND ACHIEVE CONSISTENCY WITH 20 BARN CD-111 THERMAL 4800 1451 25 CAPTURE CROSS SECTION (BALDOCK ETAL 1966 ORNL-3994) 4800 1451 26 ABOVE THE RESOLVED RESONANCE RANGE MAINLY DRAKE GA-6997 WAS USED4800 1451 27 4800 1451 28 PLEASE REFER COMMENTS OR QUESTIONS ABOUT THE TRANSLATION 4800 1451 29 CODE (UKE*** ORNL-TM-2880 OR ENDF-134) TO 4800 1451 30 R. Q. WRIGHT, OAK RIDGE NATIONAL LABORATORY, 4800 1451 31 P. O. BOX X, OAK RIDGE, TENNESSEE 37830. 4800 1451 32 4800 1451 33 RANGE EXTENDED FROM 1.0E-5 EV TO 20 MEV AND ELASTIC SCATTERING 4800 1451 34 TRANSFER AND TRANSFORMATION DATA ADDED BY NNCSC AT BNL 4800 1451 35 0 0 0 4800 1451 36 ***************** PROGRAM LINEAR (VERSION 89-1) *****************4800 1451 37 DATA LINEARIZED TO WITHIN AN ACCURACY OF 0.300 PER-CENT 4800 1451 38 ***************** PROGRAM FIXUP (VERSION 89-2) ******************4800 1451 39 *RECONSTRUCTED MT NUMBERS 4800 1451 40 4 =+( 50, 91) 4800 1451 41 103 =+(600,649) 4800 1451 42 104 =+(650,699) 4800 1451 43 105 =+(700,749) 4800 1451 44 106 =+(750,799) 4800 1451 45 107 =+(800,849) 4800 1451 46 101 =+( 18, 18)+(102,116) 4800 1451 47 27 =+(101,101) 4800 1451 48 3 =+( 4, 5)+( 16, 17)+( 22, 37)+( 41, 42) 4800 1451 49 19 =+( 18, 18)-( 20, 21)-( 38, 38) 4800 1451 50 1 =+( 2, 3) 4800 1451 51 *****************************************************************4800 1451 52 4800 1451 53 4800 1451 54 4800 1451 55 4800 1451 56 ***************** Program LINEAR (VERSION 2002-1) ***************4800 1451 57 For All Data Greater than 1.0000E-10 barns in Absolute Value 4800 1451 58 Data Linearized to Within an Accuracy of .100000000 per-cent 4800 1451 59 ***************** Program SIGMA1 (VERSION 2002-1) ***************4800 1451 60 Data Doppler Broadened to 300.000000 Kelvin 4800 1451 61 for All Data Greater than 1.0000E-10 barns in Absolute Value 4800 1451 62 Data Linearized to Within an Accuracy pf .100000000 per-cent 4800 1451 63 ***************** Program FIXUP (Version 2002-1) ****************4800 1451 64 Corrected ZA/AWR in All Sections-----------------------------Yes 4800 1451 65 Corrected Thresholds-----------------------------------------Yes 4800 1451 66 Extended Cross Sections to 20 MeV----------------------------No 4800 1451 67 Allow Cross Section Deletion---------------------------------No 4800 1451 68 Allow Cross Section Reconstruction---------------------------No 4800 1451 69 Make All Cross Sections Non-Negative-------------------------Yes 4800 1451 70 Delete Energies Not in Ascending Order-----------------------Yes 4800 1451 71 Deleted Duplicate Points-------------------------------------Yes 4800 1451 72 Check for Ascending MAT/MF/MT Order--------------------------Yes 4800 1451 73 Check for Legal MF/MT Numbers--------------------------------Yes 4800 1451 74 Allow Creation of Missing Sections---------------------------No 4800 1451 75 Allow Insertion of Energy Points-----------------------------No 4800 1451 76 Create Uniform Energy Grid-----------------------------------No 4800 1451 77 Delete Section if Cross Section =0 at All Energies-----------Yes 4800 1451 78 ***************** Program GROUPIE (VERSION 2002-1) **************4800 1451 79 Unshielded Group Averages Using 640 Groups 4800 1451 80 Weighting Spectrum: Flat (Constant) Spectrum 4800 1451 81 1 451 85 04800 1451 82 2 151 4 04800 1451 83 3 1 217 04800 1451 84 3 102 217 04800 1451 85 4800 1 099999 4800 0 0 0 4.80000E+4 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3.00000E+2 0.0 1 0 403 114931 1451 4 49-In-115 FEI/CNDC EVAL-Mar03 Zolotarev,Dunjiu,Schmittroth 4931 1451 5 DIST-Feb2004 4931 1451 6 ----IRDF-2002 MATERIAL 4931 4931 1451 7 -----INCIDENT NEUTRON DATA 4931 1451 8 ------ENDF-6 FORMAT 4931 1451 9 ***************************************************************** 4931 1451 10 49-In-115 FEI EVAL-Mar03 K.I.Zolotarev 4931 1451 11 DIST-Apr03 4931 1451 12 ----BROND-2 MATERIAL 4931 4931 1451 13 -----INCIDENT NEUTRON DATA 4931 1451 14 ------ENDF-6 FORMAT 4931 1451 15 ------Russian Reactor Dosimetry File RRDF-2002 4931 1451 16 ***************************************************************** 4931 1451 17 Authors of evaluation: K.I.Zolotarev and P.K.Zolotarev 4931 1451 18 ***************************************************************** 4931 1451 19 MF=3 4931 1451 20 MT= 51 - In115(n,n')In115m reaction 4931 1451 21 ------------------------------------- 4931 1451 22 *****************************************************************4931 1451 23 CHINESE EVALUATION OF (N,2N) REACTION ADDED AT NDS 4931 1451 24 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 4931 1451 25 4931 1451 26 THE FOLLOWING FILE SECTIONS ARE INCLUDED - 4931 1451 27 MF= 1 GENERAL INFORMATION (MT=451) 4931 1451 28 MF= 8 RADIOACTIVITY DATA (MT=51) 4931 1451 29 MF=10 CROSS SECTION FOR RADIOACTIVE NUCLIDE 4931 1451 30 PRODUCTION (MT=51) 4931 1451 31 MF=40 DATA COVARIANCES FOR RADIOACTIVE NUCLIDE 4931 1451 32 PRODUCTION (MT=51) 4931 1451 33 49-IN-115 CNDC/CIAE EVAL-JUN91 CAI DUNJIU AND WANG ZISHENG 4931 1451 34 ---ENDF-6 FORMAT 4931 REVISION 0 4931 1451 35 4931 1451 36 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 4931 1451 37 *****************************************************************4931 1451 38 49-IN-115 HEDL,ANL EVAL-MAR90 F.SCHMITTROTH,D.L.SMITH,S.CHIBA 4931 1451 39 DIST-JUN90 19900329 4931 1451 40 ----ENDF/B-VI MATERIAL 4931 4931 1451 41 MF=2 MT=151 EVALUATION OF RESOLVED RESONANCE PARAMETERS 4931 1451 42 BASED ON NEW BNL-325, REF(2). 4931 1451 43 MF=3 MT=102 VERSION-V UNRESOLVED REGION CONTAINS ADJUSTED DATA 4931 1451 44 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************4931 1451 45 DOSIMETRY 4931 1451 46 ******************************************************************4931 1451 47 ***************************************************************** 4931 1451 48 ********** Start of (N,2N) bibliographical component ********** 4931 1451 49 ***************************************************************** 4931 1451 50 4931 1451 51 Evaluation of the 115-In(n,2n) 114m-In Reaction 4931 1451 52 Cross-section 4931 1451 53 Cai Dunjiu Yu Baosheng Wang Zisheng 4931 1451 54 Lu Hanlin Zhao Wenrong 4931 1451 55 Institute of Atomic Energy, China 4931 1451 56 The excitation function of 115-In(n, 2n) 114m-In reaction 4931 1451 57 covers neutron energy range from threshold of 9.228 MeV 4931 1451 58 ( including the energy of isomeric state which has 4931 1451 59 0.190 MeV ) to 20 MeV. Experimental knowledge of the 4931 1451 60 115-In(n,2n) reaction cross sections is entirely based 4931 1451 61 on the results of activation measurements. There are 4931 1451 62 still no experimental results obtained by 4931 1451 63 direct-neutron-detection methods (such as scintillation 4931 1451 64 tank). For both the indium isotopes (115-In,0.957, 4931 1451 65 113-In,0.043), the primary activity 4931 1451 66 resulting from the (n,2n) process is due to the decay of 4931 1451 67 metastable state. The interaction with the primary isotope, 4931 1451 68 115-In, results in a metastable state in 114-In which 4931 1451 69 decays with 49.5d half-life by means of an E4 transition 4931 1451 70 to the ground state (1+) with a half-life of 71.9s. 4931 1451 71 The (n,2n) process also directly populates the ground state. 4931 1451 72 Seventeen sets of data available (1-18) were collected. 4931 1451 73 For the sake of comparison, all the experimental results were 4931 1451 74 adjusted by the unified nuclear parameters (including 4931 1451 75 intensity of r-line, half-life, branching ratio), 4931 1451 76 the reference cross section and the dependence of cross 4931 1451 77 section on energy which were taken from (Ref.1). In several 4931 1451 78 cases, adjustments were made only for the dependence of cross 4931 1451 79 section on energy since some necessary parameters used in data 4931 1451 80 analysis were not given by the authors. 4931 1451 81 The evaluation was made as the following ways: 4931 1451 82 (1).At 14.7MeV neutron energy . 4931 1451 83 Based on the results measured by Ge(Li) detector,the 4931 1451 84 recommended cross section at 14.7 MeV is 1290 mb. 4931 1451 85 (2). For the neutron energy range from threshold to 20 MeV. 4931 1451 86 The recommended excitation function was obtained 4931 1451 87 by fitting the measured data with least-square method in the 4931 1451 88 neutron energy range of 10.4-20MeV and by extrapolating from 4931 1451 89 10.4MeV to 9.228MeV based on the physical trend and systematics 4931 1451 90 calculation. 4931 1451 91 Covariance matrix is given for the evaluated data considering all 4931 1451 92 adjusted factors. 4931 1451 93 References 4931 1451 94 1.Zhao Wenrong et al., CNDC-89014;INDC(CPR)-16, 1989 4931 1451 95 2.Ke Wei et al., Chin.J.Nucl.Phys.11, 3, 11(1989) 4931 1451 96 Lu Hanlin et al., INDC(CPR)-011/GI, P67(1988) 4931 1451 97 3.J.Csikai et al., INDC(NDS)-232/L, P45(1989) 4931 1451 98 4.Li Jiangwei et al., Chin.J.Nucl.Phys., 10, 1, 52(1988) 4931 1451 99 Wu Zhihua et al., 88MITO, P315(1988) 4931 1451 100 5.T.B.Ryves et al., J.Phys., G9, 1549(1983) 4931 1451 101 6.T.B.Ryves et al., J.Phys., G7, 115(1981) 4931 1451 102 7.A.Reggoug et al., 82Antwerp, P873(1982);NIMA, 227, 249(1984) 4931 1451 103 8.K.Kayashima et al., NEANDC(J)-61U, P94(1979) 4931 1451 104 9.D.C.Santry et al., Can.J.Phys., 54, 757(1976) 4931 1451 105 10.Lu Hanlin et al., At.Energy Sci.Technol., 2, 113(1975) 4931 1451 106 11.A.Paulsen et al., Atomkernenergie, 26, 34(1975) 4931 1451 107 12.J.K.Temperley et al., BRL-1491(1970) 4931 1451 108 13.R.C.Barrall et al., Nucl.Phys., A138, 387(1969) 4931 1451 109 14.R.C.Barrall et al., AFWL-TR-68-134(1969) 4931 1451 110 15.H.Rotzer, Nucl.Phys., A109, 694(1968);DWA176, 289(1968) 4931 1451 111 16.B.Minetti et al., Z.Physik., A217, 83(1968) 4931 1451 112 17.H.O.Menlove et al., Phys.Rev., 163, 1308(1967) 4931 1451 113 18.R.J.Prestwood et al., Phys.Rev., 121, 1438(1961) 4931 1451 114 19.T.Kozlowski et al., Acta.Phys.Pol., 33, 409(1968) 4931 1451 115 20.R.Prasad et al.,Nucl.Phys., A94, 476(1967) 4931 1451 116 and 88,349(1966). 4931 1451 117 21.W.Nagel, JNE, 20, 475(1966) 4931 1451 118 22.K.C.Garg et al.,IPA17, 525(1979) 4931 1451 119 23.G.N.Salaita et al., ANS, 16, 59(1973) 4931 1451 120 24.J.Janczyszyn et al., J.Radiational Chem., 14, 201(1973) 4931 1451 121 25.M.Bormann et al., EANDC(E)76u, 51(1967) 4931 1451 122 26.W.Grochulski et al., INR-1172(1970);INR-1197, 10(1970) 4931 1451 123 Acta.Phys.Pol., BI, 271(1970) 4931 1451 124 27.H.K.Vonach et al., 68WASH, 2, P885(1968) 4931 1451 125 28.Wang Zisheng et al., Research on the Covariances Matrix 4931 1451 126 of the Evaluation Data(to be published) 4931 1451 127 4931 1451 128 ***************************************************************** 4931 1451 129 ********** End of (N,2N) bibliographical component ********** 4931 1451 130 ================================================================= 4931 1451 131 4931 1451 132 ***************************************************************** 4931 1451 133 ********** Start of (N,G) bibliographical component ********** 4931 1451 134 ***************************************************************** 4931 1451 135 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 4931 1451 136 * * * * * * *4931 1451 137 CAPTURE TO 54 MIN. ISOMERIC STATE DESCRIPTION MODIFIED BY 4931 1451 138 R. SCHENTER ON 2/20/84. THIS WAS DONE BY ADDING FILE9 WITH 4931 1451 139 Y=.79 AND PUTTING THE TOTAL CAPTURE WIDTH IN FILE2 AND THE TOTAL 4931 1451 140 CAPTURE CROSS SECTION IN FILE3. 4931 1451 141 P.C.TO NORMALIZATION+STANDARDS SUBCOMITTEE (N,G)F.SCHMITTROTH 4931 1451 142 MF=2 MT=151 4931 1451 143 MF=2 MT=102 4931 1451 144 * * * * * 4931 1451 145 4931 1451 146 FILE INFORMATION 4931 1451 147 4931 1451 148 MF=1 MT=451 ATOMIC MASS FROM REF(1). 4931 1451 149 4931 1451 150 MF=2 MT=151 EVALUATION OF RESOLVED RESONANCE PARAMETERS 4931 1451 151 BASED ON NEW BNL-325, REF(2). 4931 1451 152 MF=3 MT=102 VERSION-V UNRESOLVED REGION CONTAINS ADJUSTED DATA 4931 1451 153 SEE DOCUMENTATION 4931 1451 154 RADIATIVE NEUTRON CAPTURE TO THE IN 116M(54 MIN.) 4931 1451 155 STATE. 4931 1451 156 E GREATER THAN EH, 4931 1451 157 EVALUATION BASED ON EXPERIMENTAL DATA, REF.(3-7) 4931 1451 158 AND THEORETICAL CALCULATIONS, REF.(8,9). 4931 1451 159 E LESS THAT EH, 4931 1451 160 A 1/V COMPONENT WAS ADDED TO GIVE THE CORRECT 4931 1451 161 2200M/S CROSS SECTION TO THE 54 MIN. STATE(THE 4931 1451 162 2.2 SEC. STATE CROSS SECTION WAS INCLUDED). 4931 1451 163 RADIATIVE CAPTURE TO THE 2.2 SEC. STATE OF IN 116 WAS 4931 1451 164 INCLUDED AS PART OF THE CAPTURE TO THE 54 MIN. STATE 4931 1451 165 FOR BOTH THERMAL AND FAST ENERGIES. 4931 1451 166 RESULTS WERE DIVIDED BY .79 TO GIVE THE TOTAL CAP- 4931 1451 167 TURE CROSS SECTION IN FILE3. FILE9 COMBINED WITH 4931 1451 168 FILE3 IS REQUIRED TO GIVE CAPTURE TO 54 MIN. 4931 1451 169 ISOMERIC STATE. 4931 1451 170 4931 1451 171 2200M/S CAPTURE CROSS SECTION,BARNS TO THE 54 MIN. STATE 4931 1451 172 (FROM RESONANCE PARAMETERS) = 166.413 4931 1451 173 4931 1451 174 COMPUTED RESONANCE INTEGRAL =2587.345 4931 1451 175 4931 1451 176 REFERENCES. 4931 1451 177 1. A.H.WAPSTRA AND N.B.GOVE,NUCL. DATA TABLES,VOL.9,PART 1(1971).4931 1451 178 2. S.F.MUGHABGHAB AND D.I.GARBER,BNL-325,3RD ED.,VOL.1(1973) 4931 1451 179 3. H.A.GRENCH AND H.O.MENLOVE,PHYS.REV.,VOL.165(1968)1298. 4931 1451 180 4. H.O.MENLOVE,ET AL.,PHYS.REV.,VOL.163(1967)1299. 4931 1451 181 5. S.A.COX, PHYS.REV.,VOL.133(1964)B378. 4931 1451 182 6. A.E.JOHNSRUD,ET AL.,PHYS.REV.,VOL.116(1959)927. 4931 1451 183 7. G.PETO ET AL., J.NUCL.EN.,VOL.21(1967)797. 4931 1451 184 8. F.SCHMITTROTH, HEDL-TME 71-106(AUGUST 1971). 4931 1451 185 9. F.SCHMITTROTH, HEDL-TME 73-79,ENDF-195(NOVEMBER 1973). 4931 1451 186 ***************************************************************** 4931 1451 187 ********** (N,G) processing details ********** 4931 1451 188 ***************************************************************** 4931 1451 189 Data Linearized to Within an Accuracy of .100000000 per-cent 4931 1451 190 ***************** Program RECENT (VERSION 2002-1) ***************4931 1451 191 for All Data Greater than 1.0000E-10 barns in Absolute Value 4931 1451 192 Data Linearized to within an Accuracy of .100000000 per-cent 4931 1451 193 ***************** Program SIGMA1 (VERSION 2002-1) ***************4931 1451 194 Data Doppler Broadened to 300.000000 Kelvin 4931 1451 195 for All Data Greater than 1.0000E-10 barns in Absolute Value 4931 1451 196 Data Linearized to Within an Accuracy pf .100000000 per-cent 4931 1451 197 ***************** Program FIXUP (Version 2002-1) ****************4931 1451 198 Corrected ZA/AWR in All Sections-----------------------------Yes 4931 1451 199 Corrected Thresholds-----------------------------------------Yes 4931 1451 200 Extended Cross Sections to 20 MeV----------------------------No 4931 1451 201 Allow Cross Section Deletion---------------------------------No 4931 1451 202 Allow Cross Section Reconstruction---------------------------No 4931 1451 203 Make All Cross Sections Non-Negative-------------------------Yes 4931 1451 204 Delete Energies Not in Ascending Order-----------------------Yes 4931 1451 205 Deleted Duplicate Points-------------------------------------Yes 4931 1451 206 Check for Ascending MAT/MF/MT Order--------------------------Yes 4931 1451 207 Check for Legal MF/MT Numbers--------------------------------Yes 4931 1451 208 Allow Creation of Missing Sections---------------------------No 4931 1451 209 Allow Insertion of Energy Points-----------------------------No 4931 1451 210 Create Uniform Energy Grid-----------------------------------No 4931 1451 211 Delete Section if Cross Section =0 at All Energies-----------Yes 4931 1451 212 *************** PROGRAM ACTIVATE (VERSION 2002-1) ***************4931 1451 213 MF=10 Activation Cross Sections Defined by Combining MF=3 4931 1451 214 Cross Sections and MF=9 Multipliers. MF=9 Deleted. 4931 1451 215 ***************************************************************** 4931 1451 216 The metastable state component (MF/MT=10/102) was extracted from 4931 1451 217 the ACTIVATE output and merged with the other reactions in this 4931 1451 218 In-115 evaluation. MF/MT=33/102 was adapted as MF/MT=40/102. 4931 1451 219 ***************************************************************** 4931 1451 220 ********** End of (N,G) bibliographical component ********** 4931 1451 221 ================================================================= 4931 1451 222 ********** Start of (N,N') bibliographical component ********** 4931 1451 223 ***************************************************************** 4931 1451 224 ------Russian Reactor Dosimetry File RRDF-2002 4931 1451 225 ***************************************************************** 4931 1451 226 Authors of evaluation: K.I.Zolotarev and P.K.Zolotarev 4931 1451 227 ***************************************************************** 4931 1451 228 MF=3 4931 1451 229 MT= 51 - In115(n,n')In115m reaction 4931 1451 230 ------------------------------------- 4931 1451 231 In the section MT= 51 are given evaluated cross section data 4931 1451 232 of neutron inelastic excitation of the 4.486-hour isomeric state 4931 1451 233 in the indium-115 in the energy range from threshold to 20 MeV. 4931 1451 234 The isomeric level has energy 336.24 keV with spin and parity 4931 1451 235 1/2- [1]. 4931 1451 236 Microscopic experimental data [2-40] were analyzed in the 4931 1451 237 process of preparation of input data base for the evaluation of 4931 1451 238 cross sections and their uncertainty for the In-115(n,n')In-115m 4931 1451 239 reaction. During this procedure all experimental data if it was 4931 1451 240 possible were corrected to the new recommended cross section data 4931 1451 241 for monitor reactions used in the measurements and to the new re- 4931 1451 242 commended decay data. 4931 1451 243 Special correction was done with experimental data [5], [6], 4931 1451 244 [12], [24], [25] and [27]. 4931 1451 245 Grench and Menlove data [5] and Kimura et al. data [6] bet- 4931 1451 246 ween 0.5 - 0.9 MeV obtained in the measurements with T(p,n)He3 4931 1451 247 neutron source were renormalized to the integral of Liskien et al.4931 1451 248 experimental data [16] in the overlapping energy ranges. Original 4931 1451 249 cross section data from ref. [5] and [6] were multiplied to the 4931 1451 250 correction factors Fc=1.15268 and Fc=0.92486 , respectively. 4931 1451 251 Data of Santry and Butler [12] obtained in the absolute measu-4931 1451 252 rements with T(p,n)He3 source in the energy range 0.348-5.130 MeV 4931 1451 253 were renormalized to the integral of Liskien et al. experimental 4931 1451 254 data [16] in the energy interval 1.85-4.07 MeV, Fc= 1.0516 4931 1451 255 Cross section data measured by Zhao Wenrong et al. [24] were 4931 1451 256 renormalized also to the experimental data of Liskien et al. [16] 4931 1451 257 in the overlapping energy ranges. All original experimental data 4931 1451 258 from ref. [24] except cross section value at the energy point 4931 1451 259 En=8.5 MeV were multiplied to the factor Fc=1.065 . 4931 1451 260 Experimental data of Kimura et al. [6] in the energy interval 4931 1451 261 1.8-4.6 MeV and experimental data of Santry and Butler [12] obta- 4931 1451 262 ined relative to the S-32(n,p)P-32 standard were not corrected. 4931 1451 263 Data of Zhao Wenrong et al. [24] for the energy point 8.5 MeV 4931 1451 264 was corrected to the recommended cross section value for monitor 4931 1451 265 tor reaction Al-27(n,a)Na-24 from ref. [41]. 4931 1451 266 Cross section data for the In-115(n,n')In-115m reaction measu-4931 1451 267 red by Lu Hanlin et al. [25] and by Csikai et al. [27] were renor-4931 1451 268 malized to the integral of Konno et al. [30] and Filatenkov et al.4931 1451 269 [31] experimental data in the overlapping energy interval 13.54 - 4931 1451 270 14.78 MeV . Original data from ref. [25] and [27] were multiplied 4931 1451 271 to the correction factors Fc=1.13622 and Fc=1.08624, respectively.4931 1451 272 Excitation function for the In-115(n,n')In-115m reaction in 4931 1451 273 the energy region from threshold to 20 MeV was evaluated by means 4931 1451 274 of statistical analysis of experimental cross section data [2-31].4931 1451 275 Experimental cross section data [32-40] were rejected due to 4931 1451 276 their discrepancy with the main bulk of experimental data [2-31]. 4931 1451 277 Data of Martin et al. [2] in the energy interval 4.55 - 5.26 MeV 4931 1451 278 were not taken into account in the evaluation due to their syste- 4931 1451 279 matic underestimation In-115(n,n')In-115m reaction cross section. 4931 1451 280 In the rejected experiments [33-38] the cross section values were 4931 1451 281 measured only in a one energy point in the interval 14 - 15 MeV. 4931 1451 282 Statistical analysis of input cross section data was carried 4931 1451 283 out by means of PADE-2 code [42]. Rational function was used as 4931 1451 284 the model function [43]. 4931 1451 285 U-235 thermal fission [44] and Cf-252 spontaneous fission 4931 1451 286 neutron spectra [45] averaged cross sections calculated from the 4931 1451 287 the evaluated In-115(n,n')In-115m excitation function are the 4931 1451 288 following: 4931 1451 289 4931 1451 290 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4931 1451 291 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 4931 1451 292 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4931 1451 293 U-235 neutron fission ³ 188.40 ³ 188.2 +- 2.3 [46] 4931 1451 294 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4931 1451 295 Cf-252 spont. fission ³ 191.66 ³ 197.4 +- 2.704 [47] 4931 1451 296 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 4931 1451 297 4931 1451 298 MT=33 4931 1451 299 MT= 51 -(n,n') cross section cov. matrix 4931 1451 300 ---------------------------------------- 4931 1451 301 Uncertainties in the evaluated excitation function for the 4931 1451 302 reaction In-115(n,n')In-115m are given in the form of relative 4931 1451 303 covariance matrix for the 39-neutron energy groups (LB=5). Cova- 4931 1451 304 riance matrix of uncertainties was calculated simultaneously with 4931 1451 305 recommended cross section data by means of PADE-2 code. 4931 1451 306 Eigenvalues of the 6-th digits relative covariance matrix 4931 1451 307 given in the 33-file are the following: 4931 1451 308 4931 1451 309 7.55059E-08 8.01529E-08 8.14328E-08 8.29955E-08 4931 1451 310 8.51539E-08 8.79248E-08 8.88394E-08 9.29091E-08 4931 1451 311 9.80564E-08 1.06991E-07 1.19646E-07 1.34061E-07 4931 1451 312 1.46033E-07 1.71437E-07 2.22942E-07 2.66525E-07 4931 1451 313 4.04740E-07 7.05147E-07 1.52953E-06 2.70059E-06 4931 1451 314 2.95312E-06 3.08776E-06 6.35038E-06 4.26321E-04 4931 1451 315 6.81158E-04 9.21024E-04 9.54240E-04 1.07694E-03 4931 1451 316 1.11466E-03 1.28566E-03 1.45684E-03 1.62008E-03 4931 1451 317 1.97248E-03 4.10930E-03 5.75040E-03 7.29359E-03 4931 1451 318 1.00269E-02 1.64825E-02 2.00540E-02 4931 1451 319 4931 1451 320 References: 4931 1451 321 1. R.B.Firestone Table of Isotopes, Eighth edition, Vol. 1, 4931 1451 322 John Wiley & Sons, Inc., New York, 1995 4931 1451 323 2. H.C.Martin et al. Phys. Rev., v.93, p.199, January 1954 4931 1451 324 3. H.O.Menlove, K.L.Coop, H.A.Grench, R.Sher Phys. Rev., v.163, 4931 1451 325 p.1299, 1967 4931 1451 326 4. P.Bornemisza-Pauspertl et al. Atomki Koezlemenyek, v.10, No.2,4931 1451 327 p.112, July 1968 4931 1451 328 5. H.A.Grench, H.O.Menlove Physical Review, v.165, p.1298, 1968 4931 1451 329 6. I.Kimura et al. Journal of Nucl. Science and Technology, v.6,4931 1451 330 p.485, September 1969 4931 1451 331 7. R.C.Barrall et al. Report AFWL-TR-68-134, Albuquerque, NM, 4931 1451 332 March 1969 ; 4931 1451 333 R.C.Barrall et al. Nucl. Phys. A, v.138, p.387, December 1969 4931 1451 334 8. J.K.Temperley, D.E.Barnes Report BRL-R-1491, August 1970 4931 1451 335 9. L.Lakosi, A.Veres 2nd Nat'l Soviet Conf. on Neutron Physics, 4931 1451 336 Kiev, v.4, p.312, May 1973 4931 1451 337 10. K.Kobayashi et al. J. Nucl. Energ., v.27, p.741, October 1973 4931 1451 338 11. D.L.Smith, J.W.Meadows Report ANL-NDM-14, Argonne, July 1975; 4931 1451 339 D.L.Smith, J.W.Meadows Nucl. Sci. Eng., v.60, p.319,July 1976 4931 1451 340 12. D.C.Santry, J.P.Butler Can. J. Phys., v.54, p.757, 1976 4931 1451 341 13. C.G.Hudson, W.L.Alford Bull. Amer. Phys. Soc., v.21, 4931 1451 342 p.188(DB5), February 1976 4931 1451 343 14. G.Magnusson, I.Bergqvist Nuclear Technology, v.34, p.114, 4931 1451 344 June 1977 4931 1451 345 15. S.Yamamoto et al. Report NEANDC(J)-56, p.25, September 1978 4931 1451 346 16. H.Liskien et al. Nucl. Sci. Eng., v.67, p.334, September 1978 4931 1451 347 17. P.Andersson et al. Report LUNF-D6-3021, University of Lund, 4931 1451 348 November 1978 4931 1451 349 18. C.F.Ai, J.C.Chou Nuclear Science, Taiwan, v.16, no.3, p.157, 4931 1451 350 September 1979 4931 1451 351 19. L.Adamski et al. Annals of Nucl. Energy, v.7, p.397, 1980 4931 1451 352 20. Fan Pei-Guo et al. Chinese J. of Nuclear Physics, v.2, No.4, 4931 1451 353 p.337, November 1980 4931 1451 354 21. V.L.Demekhin et al. 6th All Union Conference on Neutron 4931 1451 355 Physics, Kiev, 2 - 6 October, v.3, p.195, 1983 4931 1451 356 22. T.B.Ryves et al. J. of Physics, pt.G, v.9, p.1549, Dec. 1983 4931 1451 357 23. K.Kudo et al. Report NEANDC(J)-106/U, p.1, September 1984 4931 1451 358 24. Zhao Wenrong, Lu Hanlin et al. Report INDC(CPR)-16, Sep. 1989 4931 1451 359 25. Lu Hanlin et al. Chinese J. of Nuclear Physics, v.11, n.2, 4931 1451 360 p.53, 1989 4931 1451 361 26. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 4931 1451 362 27. J.Csikai et al. Zeitschrift fuer Physik, v.A337, p.39, 1990 4931 1451 363 28. N.N.Moiseev et al. Atomnaja Energija, v.68, February 1990 4931 1451 364 29. Y.Ikeda et al. Proc. of an Int. Conf. Nuclear Data for Sci- 4931 1451 365 ence and Technology, Julich, FRG, 13-17 May 1991, Springer- 4931 1451 366 Verlag, 1992, pp.294-296 4931 1451 367 30. C.Konno et al. Report JAERI-1329, October 1993 4931 1451 368 31. A.A.Filatenkov et al. Report RI-252, St. Petesburg, May 1999; 4931 1451 369 A.A.Filatenkov et al. VANT, Ser.:Yadernye Konstanty, v.2, p.8,4931 1451 370 Moscow, 1996 4931 1451 371 32. S.G.Cohen Nature, v.161, p.475, March 1948 4931 1451 372 33. W.Nagel EXFOR 20198.016 , December 1966 4931 1451 373 34. H.Roetzer Nucl. Phys., v.A109, p.694, March 1968 4931 1451 374 35. B.Minetti, A.Pasquarelli Zeitschrift f. Physik, v.217, p.83, 4931 1451 375 October 1968 4931 1451 376 36. I.Garlea et al. Rev. Roum. Phys., v.29, p.421, 1984 4931 1451 377 37. R.Pepelnik et al. Report NEANDC(E)-262U, No.5,p.32, June 1985 4931 1451 378 38. I.Garlea et al. Rev. Roum. Phys., v.37, no.1, pp.19-25, 1992 4931 1451 379 39. J.Csikai, P.Raics EXFOR 30983.002 July 1992 4931 1451 380 40. N.I.Molla et al. EXFOR 30984.002 4931 1451 381 41. H.Vonach in "Nuclear Data Standards for Nuclear Measurements",4931 1451 382 Report NEANDC-311 U, pp.75-77, OECD, Paris, 1992. ; 4931 1451 383 M.Wagner et al. Evaluation of cross sections for 14 important 4931 1451 384 neutron-dosimetry reactions. Physics Data No.13-5, 1990 4931 1451 385 42. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 4931 1451 386 43. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 4931 1451 387 st's Meeting on Evaluation and Processing of Covariance Data, 4931 1451 388 Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105 4931 1451 389 44. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 4931 1451 390 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 4931 1451 391 45. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 4931 1451 392 46. W.Mannhart Prog. Report INDC(Ger)-045, pp.40-43, June 1999 4931 1451 393 47. W.Mannhart Validation of Differential Cross Sections with 4931 1451 394 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 4931 1451 395 September 2002 4931 1451 396 ***************************************************************** 4931 1451 397 ********** End of (N,N') bibliographical component ********** 4931 1451 398 ================================================================= 4931 1451 399 4931 1451 400 4931 1451 401 4931 1451 402 4931 1451 403 ***************************************************************** 4931 1451 404 ***************** Program GROUPIE (VERSION 2002-1) **************4931 1451 405 Unshielded Group Averages Using 640 Groups 4931 1451 406 Weighting Spectrum: Flat (Constant) Spectrum 4931 1451 407 1 451 418 14931 1451 408 2 151 4 14931 1451 409 8 4 2 14931 1451 410 8 16 2 14931 1451 411 8 102 2 14931 1451 412 10 4 75 14931 1451 413 10 16 40 14931 1451 414 10 102 217 14931 1451 415 40 4 148 14931 1451 416 40 16 24 14931 1451 417 40 102 7 04931 1451 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3.634430-5493140 4 142 1.177520-3 1.652270-3 1.220020-3 2.789710-4-8.913250-4 8.698960-4493140 4 143 3.475700-4 3.534930-4 4.186150-4 3.550390-4 1.773520-4-7.331360-5493140 4 144 5.153350-4 3.975890-4 2.332660-4 1.091980-4 2.823100-5-2.163150-5493140 4 145 1.382620-3 1.844060-3 1.642420-3 1.011630-3 1.406180-4 2.928410-3493140 4 146 3.028540-3 2.435580-3 1.422320-3 3.697970-3 3.780810-3 3.486050-3493140 4 147 4.941200-3 5.924580-3 8.514690-3 493140 4 148 493140 099999 4.91150E+4 1.13917E+2 0 0 1 0493140 16 1 -9.035000+6-9.225000+6 0 0 0 1493140 16 2 1.000000+1 0.000000+0 0 16 0 1493140 16 3 0.000000+0 0.000000+0 1 5 120 15493140 16 4 1.000000-5 9.100000+6 1.050000+7 1.150000+7 1.220000+7 1.270000+7493140 16 5 1.320000+7 1.380000+7 1.420000+7 1.470000+7 1.520000+7 1.600000+7493140 16 6 1.720000+7 1.800000+7 2.000000+7 4.000000-1 1.154500-2 2.886700-3493140 16 7 5.207500-4 6.268700-4 2.634000-4 7.576400-5 0.000000+0 0.000000+0493140 16 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.360500-3493140 16 9 2.329300-4 4.379300-5 3.313900-5 2.375700-5 6.429400-6 0.000000+0493140 16 10 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0493140 16 11 4.273200-4 5.149000-5 5.671700-5 3.970200-5 1.611800-5 0.000000+0493140 16 12 5.134000-6 5.330100-6 0.000000+0 3.850900-6 2.755800-7 0.000000+0493140 16 13 1.326100-4 7.026400-5 5.057000-5 1.814000-5 1.025800-5 1.091900-5493140 16 14 1.151900-5 6.054300-6 4.032200-6 3.183000-6 2.577000-7 1.061600-4493140 16 15 7.193900-5 1.827700-5 7.806100-6 1.063900-5 1.195100-5 9.623500-6493140 16 16 6.619700-6 5.063800-6 1.300400-6 9.380800-5 2.002300-5 8.788300-6493140 16 17 1.212000-5 1.349000-5 1.062800-5 7.704500-6 6.688400-6 1.473800-6493140 16 18 5.145600-5 2.270400-5 1.436100-5 1.686000-5 4.715200-6 4.452700-6493140 16 19 4.233200-6 1.200700-6 4.804900-5 1.232200-5 1.347800-5 3.765900-6493140 16 20 2.757000-6 3.388400-6 4.405400-7 4.707100-5 2.134000-5 8.565900-6493140 16 21 1.578500-5 9.719500-6 9.119700-6 4.580100-5 1.385300-5 1.154500-5493140 16 22 1.754500-5 3.156100-6 4.554700-5 1.336600-5 2.469400-5 2.724500-6493140 16 23 4.622500-5 2.217600-5 1.374800-5 7.898800-5 5.926800-6 8.012400-5493140 16 24 493140 099999 4.91150E+4 1.13917E+2 0 0 1 0493140102 1 6.598000+6 6.471000+6 0 1 0 1493140102 2 1.000000+1 1.000000+0 0 102 0 1493140102 3 0.000000+0 0.000000+0 0 1 14 7493140102 4 1.000000-5 3.600000-3 1.000000+3 2.250000-2 1.000000+4 8.100000-3493140102 5 1.000000+5 2.500000-3 1.000000+6 4.900000-3 4.000000+6 1.690000-2493140102 6 2.000000+7 0.000000+0 493140102 7 493140 099999 4931 0 0 0 0 0 0 0 5.31270E+4 1.25814E+2 0 0 34 105325 1451 1 0.0 0.0 0 0 0 65325 1451 2 1.00000E+0 2.00000E+7 0 0 10 20025325 1451 3 3.00000E+2 0.0 1 0 60 45325 1451 4 53-I -127 CNDC/CIAE EVAL-JUN91 Zhao Wenrong and Lu Hanlin et.al.5325 1451 5 DIST-Feb2004 5325 1451 6 ----IRDF-2002 MATERIAL 5325 5325 1451 7 -----INCIDENT NEUTRON DATA 5325 1451 8 ------ENDF-6 FORMAT 5325 1451 9 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 5325 1451 10 53-I -127 CNDC/CIAE EVAL-JUN91 Zhao Wenrong and Lu Hanlin et.al.5325 1451 11 ---ENDF-6 FORMAT 5325 REVISION 0 5325 1451 12 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 5325 1451 13 5325 1451 14 Results of the evaluation in ENDF/B Format 5325 1451 15 53-I-127 5325 1451 16 I-127(n,2n)I-126 Dosimetry reaction 5325 1451 17 Evaluated by Zhao Wenrong Lu Hanlin Yu Weixiang Cai Dunjiu 5325 1451 18 Wang Zisheng Huang Xiaolong 5325 1451 19 5325 1451 20 The reaction of I-127(n,2n)I-126 is used as a threshold detector 5325 1451 21 for neutron dosimetry applications. Its threshold is 9.137 5325 1451 22 MeV. The half-life of I-126 is 13.023d. Iodine is a monoisotopic 5325 1451 23 element. The main gamma ray energies of I-126 are 388.6keV 5325 1451 24 and 666.4keV with branch ratios of 34.1 percent and 33.1 percent 5325 1451 25 respectively. In order to get reasonable results, the necessary 5325 1451 26 adjustment was made first for the collected data. Then the 5325 1451 27 evaluation was performed by spline fitting procedure. The 5325 1451 28 uncertainties were derived from experimental errors and 5325 1451 29 the consideration of systematics. 5325 1451 30 5325 1451 31 Evaluation documentation: Zhao Wenrong. 5325 1451 32 Report CIAE 1991 5325 1451 33 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 5325 1451 34 The Q values and threshold energies were updated prior to pro- 5325 1451 35 cessing through the codes to comply with the values obtained 5325 1451 36 using the NNDC calculation program which is based on the 1995 5325 1451 37 Update to the Atomic mass Evaluation. 5325 1451 38 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * 5325 1451 39 ***************** Program LINEAR (VERSION 2002-1) ***************5325 1451 40 For All Data Greater than 1.0000E-10 barns in Absolute Value 5325 1451 41 Data Linearized to Within an Accuracy of .100000000 per-cent 5325 1451 42 ***************** Program SIGMA1 (VERSION 2002-1) ***************5325 1451 43 Data Doppler Broadened to 300.000000 Kelvin 5325 1451 44 for All Data Greater than 1.0000E-10 barns in Absolute Value 5325 1451 45 Data Linearized to Within an Accuracy pf .100000000 per-cent 5325 1451 46 ***************** Program FIXUP (Version 2002-1) ****************5325 1451 47 Corrected ZA/AWR in All Sections-----------------------------Yes 5325 1451 48 Corrected Thresholds-----------------------------------------Yes 5325 1451 49 Extended Cross Sections to 20 MeV----------------------------No 5325 1451 50 Allow Cross Section Deletion---------------------------------No 5325 1451 51 Allow Cross Section Reconstruction---------------------------No 5325 1451 52 Make All Cross Sections Non-Negative-------------------------Yes 5325 1451 53 Delete Energies Not in Ascending Order-----------------------Yes 5325 1451 54 Deleted Duplicate Points-------------------------------------Yes 5325 1451 55 Check for Ascending MAT/MF/MT Order--------------------------Yes 5325 1451 56 Check for Legal MF/MT Numbers--------------------------------Yes 5325 1451 57 Allow Creation of Missing Sections---------------------------No 5325 1451 58 Allow Insertion of Energy Points-----------------------------No 5325 1451 59 Create Uniform Energy Grid-----------------------------------No 5325 1451 60 Delete Section if Cross Section =0 at All Energies-----------Yes 5325 1451 61 ***************** Program GROUPIE (VERSION 2002-1) **************5325 1451 62 Unshielded Group Averages Using 640 Groups 5325 1451 63 Weighting Spectrum: Flat (Constant) Spectrum 5325 1451 64 1 451 68 15325 1451 65 2 151 4 15325 1451 66 3 16 40 15325 1451 67 33 16 19 15325 1451 68 5325 1 099999 5325 0 0 0 5.31270E+4 1.25814E+2 0 0 1 05325 2151 1 5.31270E+4 1.00000E+0 0 0 1 05325 2151 2 1.00000E-5 1.98900E+1 0 0 0 05325 2151 3 2.50000E+0 5.70000E-1 0 0 0 05325 2151 4 5325 2 099999 5325 0 0 0 5.31270E+4 1.25814E+2 0 0 1 05325 3 16 1 -9.14300E+6-9.14300E+6 0 0 1 1095325 3 16 2 109 1 5325 3 16 3 9200000.00 .002125030 9300000.00 .003744262 9400000.00 .0058736845325 3 16 4 9500000.00 .008981316 9600000.00 .031000000 9700000.00 .0660600005325 3 16 5 9800000.00 .108200000 9900000.00 .155240000 10000000.0 .2070000005325 3 16 6 10100000.0 .262032500 10200000.0 .320900000 10300000.0 .3824372375325 3 16 7 10400000.0 .448368421 10500000.0 .517276842 10600000.0 .5849000005325 3 16 8 10700000.0 .651900000 10800000.0 .718900000 10900000.0 .7858500005325 3 16 9 11000000.0 .851150000 11100000.0 .915570000 11200000.0 .9773500005325 3 16 10 11300000.0 1.03776632 11400000.0 1.09542105 11500000.0 1.151272635325 3 16 11 11600000.0 1.19980000 11700000.0 1.24544000 11800000.0 1.285400005325 3 16 12 11900000.0 1.32310750 12000000.0 1.35620000 12100000.0 1.387452505325 3 16 13 12200000.0 1.41480000 12300000.0 1.44060105 12400000.0 1.463263165325 3 16 14 12500000.0 1.48443579 12600000.0 1.50185000 12700000.0 1.518110005325 3 16 15 12800000.0 1.53215000 12900000.0 1.54537000 13000000.0 1.556000005325 3 16 16 13100000.0 1.56600000 13200000.0 1.57600000 13300000.0 1.585794875325 3 16 17 13400000.0 1.59369231 13500000.0 1.60112821 13600000.0 1.608564105325 3 16 18 13700000.0 1.61572676 13800000.0 1.62125000 13900000.0 1.626500005325 3 16 19 14000000.0 1.63175000 14100000.0 1.63679830 14200000.0 1.640636365325 3 16 20 14300000.0 1.64427273 14400000.0 1.64790909 14500000.0 1.651545455325 3 16 21 14600000.0 1.65518182 14700000.0 1.65826506 14800000.0 1.660795185325 3 16 22 14900000.0 1.66332530 15000000.0 1.66585542 15100000.0 1.668385545325 3 16 23 15200000.0 1.67091566 15300000.0 1.67344578 15400000.0 1.675975905325 3 16 24 15500000.0 1.67817198 15600000.0 1.67940000 15700000.0 1.680566675325 3 16 25 15800000.0 1.68173333 15900000.0 1.68290000 16000000.0 1.684066675325 3 16 26 16100000.0 1.68494750 16200000.0 1.68500000 16300000.0 1.684748725325 3 16 27 16400000.0 1.68376923 16500000.0 1.68274359 16600000.0 1.681717955325 3 16 28 16700000.0 1.68006051 16800000.0 1.67710000 16900000.0 1.674100005325 3 16 29 17000000.0 1.67110000 17100000.0 1.66762000 17200000.0 1.663150005325 3 16 30 17300000.0 1.65795667 17400000.0 1.65133333 17500000.0 1.644666675325 3 16 31 17600000.0 1.63800000 17700000.0 1.63028708 17800000.0 1.621050005325 3 16 32 17900000.0 1.61180000 18000000.0 1.60255000 18100000.0 1.592388755325 3 16 33 18200000.0 1.58090000 18300000.0 1.56824212 18400000.0 1.553897445325 3 16 34 18500000.0 1.53953846 18600000.0 1.52517949 18700000.0 1.509750005325 3 16 35 18800000.0 1.49325000 18900000.0 1.47550000 19000000.0 1.456500005325 3 16 36 19100000.0 1.43675000 19200000.0 1.41625000 19300000.0 1.393692315325 3 16 37 19400000.0 1.36907692 19500000.0 1.34446154 19600000.0 1.319880525325 3 16 38 19700000.0 1.29935484 19800000.0 1.28161290 19900000.0 1.263870975325 3 16 39 20000000.0 0.0 5325 3 16 40 5325 3 099999 5325 0 0 0 5.31270E+4 1.25814E+2 0 0 1 1532533 16 1 0.000000+0 0.000000+0 0 16 0 1532533 16 2 0.000000+0 0.000000+0 1 5 91 13532533 16 3 1.000000-5 9.200000+6 1.058000+7 1.118000+7 1.157000+7 1.237000+7532533 16 4 1.355000+7 1.415000+7 1.454000+7 1.514000+7 1.712000+7 1.811000+7532533 16 5 1.989000+7 1.000000-1 2.189400-3 1.465900-3 8.864000-4 6.104300-4532533 16 6 5.784600-4 5.628800-4 4.006100-4 5.706500-4 2.236400-4 3.653400-4532533 16 7 3.326600-4 2.851600-4 6.491500-5 3.944500-5 2.716400-5 2.613800-5532533 16 8 2.527600-5 1.805000-5 2.583600-5 9.952100-6 1.661900-5 1.522700-5532533 16 9 1.278300-4 2.641000-5 1.818700-5 1.750000-5 1.692300-5 1.208500-5532533 16 10 1.729800-5 6.663300-6 1.112700-5 1.019500-5 6.062500-5 1.953900-5532533 16 11 2.045100-5 8.819600-6 9.864000-6 7.941700-6 3.976900-6 1.965700-5532533 16 12 1.911200-5 4.140300-5 1.991900-5 1.119300-5 9.425200-6 1.161100-5532533 16 13 1.011200-5 1.294000-5 1.031500-5 2.200000-5 9.677000-6 7.798900-6532533 16 14 1.012000-5 7.187100-6 7.827200-6 6.343900-6 1.636500-5 1.249200-5532533 16 15 1.206100-5 9.854400-6 9.520400-6 8.511300-6 1.567300-5 1.273500-5532533 16 16 7.466100-6 8.300400-6 6.147800-6 1.539000-5 1.140600-5 1.040700-5532533 16 17 8.057200-6 1.543700-5 1.378500-5 8.856800-6 1.830900-5 1.393200-5532533 16 18 2.509400-5 532533 16 19 532533 099999 5325 0 0 0 0 0 0 0 5.71390E+4 1.37713E+2 0 0 34 105728 1451 1 0.0 0.0 0 0 0 65728 1451 2 1.00000E+0 2.00000E+7 0 0 10 20025728 1451 3 3.00000E+2 0.0 1 0 200 45728 1451 4 57-La-139 FEI EVAL-SEP01 K.I.Zolotarev 5728 1451 5 DIST-Feb2004 5728 1451 6 ----IRDF-2002 MATERIAL 5728 5728 1451 7 -----INCIDENT NEUTRON DATA 5728 1451 8 ------ENDF-6 FORMAT 5728 1451 9 ***************************************************************** 5728 1451 10 57-La-139 FEI EVAL-Sep01 K.I.Zolotarev 5728 1451 11 DIST-Jan04 5728 1451 12 ----BROND-3 MATERIAL 5728 Revision 1, Jan. 2004 5728 1451 13 -----INCIDENT NEUTRON DATA 5728 1451 14 ------ENDF-6 FORMAT 5728 1451 15 ***************************************************************** 5728 1451 16 ------Russian Reactor Dosimetry File RRDF-2002 5728 1451 17 ***************************************************************** 5728 1451 18 Authors of evaluation: K.I.Zolotarev and V.G.Pronyaev 5728 1451 19 ***************************************************************** 5728 1451 20 5728 1451 21 ----- MF=2 MT=151 ----- 5728 1451 22 5728 1451 23 RESONANCE PARAMETERS 5728 1451 24 1. Resolved Resonance Region (RRR) 5728 1451 25 Resolved MLBW resonance parameters up to 20 keV were based 5728 1451 26 on data from ref.[1-2]. Spins of resonances, which were not iden- 5728 1451 27 tified, have been assigned by taking into account 2*J+1 law and 5728 1451 28 approximate linear growth of the number of levels with given spin 5728 1451 29 and parity from energy. Average gamma-widths 0.055 eV (L=0) and 5728 1451 30 0.04 eV (L=1) were assigned to those resonances , widths for 5728 1451 31 which were absent. Fictitious P-resonances were added in the 5728 1451 32 energy region (3 - 20) keV to obtain the average capture cross 5728 1451 33 section values observed in the experiments [3],[4],[5]. 5728 1451 34 5728 1451 35 2. Unresolved Resonance Region (URR) 5728 1451 36 Unresolved resonance parameters in the region 20 keV - 160 keV 5728 1451 37 were evaluated by using 'EVPAR' code (Hauser-Feshbach-Moldauer 5728 1451 38 statistical model) [6]. URR parameters were calculated from opti- 5728 1451 39 mized fit to the measured capture [4,5] and total cross section 5728 1451 40 [7,8,9] data. Average resolved resonance parameters have been 5728 1451 41 used as zero approximation. Contribution of F-wave in the energy 5728 1451 42 interval 60 - 160 keV given as smoothed background cross section 5728 1451 43 in File 3. 5728 1451 44 5728 1451 45 ----- MF=3 MT=102 ----- 5728 1451 46 5728 1451 47 Capture cross sections from 1.0E-5 eV to 160 keV are reconst- 5728 1451 48 ructed from evaluated MLBW resolved and unresolved resonance pa- 5728 1451 49 rameters. Small background was added in RRR for taking into 5728 1451 50 account non-statistical reaction mechanism contribution. Contri- 5728 1451 51 bution of F-wave in the energy interval (60 - 160) keV was added 5728 1451 52 as described above. 5728 1451 53 Data base for the evaluation La-139(n,g)La-140 excitation func-5728 1451 54 tion in the neutron energy range 160 keV - 20 MeV was formed from 5728 1451 55 microscopic experimental data [4-5], [10-28] and data received 5728 1451 56 from theoretical model calculations. Experimental data [4], [10], 5728 1451 57 [12-18], [20-22], [27-28] were corrected to the new standards.5728 1451 58 Uncertainties for cross section data measured by Johnsrud et al.5728 1451 59 [13] were evaluated between +-(16-19)%. Theoretical model calcula-5728 1451 60 tions was done by means of GNASH code [29]. New modified version 5728 1451 61 of GNASH was utilized for calculations [30]. Data from theoreti-5728 1451 62 cal model calculation were dominant in the neutron energy range 5728 1451 63 3.5 - 20 MeV. 5728 1451 64 The evaluation La-139(n,g)La-140 excitation function in the 5728 1451 65 energy range above 160 keV has been carried out within the frame- 5728 1451 66 work of generalized least squares method , rational function was 5728 1451 67 used as model function [31]. Procedure of calculation recommended 5728 1451 68 cross section data was performed by means of PADE-2 code [32]. 5728 1451 69 Capture cross section at En=0.0253 eV and Resonance Integral 5728 1451 70 (0.5 to 2.0E+7 eV) from present evaluation are given below in the 5728 1451 71 Table 1. in a comparison with data from compilations [1], [33-34].5728 1451 72 Table 1. 5728 1451 73 ------------------------------------------------- 5728 1451 74 Parameters Values, barn References 5728 1451 75 ------------------------------------------------- 5728 1451 76 Capture cross section 9.04 +- 0.23 This eval. 5728 1451 77 at En=0.0253 eV 8.93 +- 0.04 [ 1] 5728 1451 78 9.04 +- 0.04 [33] 5728 1451 79 9.2 +- 0.2 [34] 5728 1451 80 ------------------------------------------------- 5728 1451 81 Resonance Integral 12.1 This eval. 5728 1451 82 11.8 +- 0.8 [ 1] 5728 1451 83 12.1 +- 0.6 [33] 5728 1451 84 11.8 +- 0.8 [34] 5728 1451 85 ------------------------------------------------- 5728 1451 86 5728 1451 87 ----- MF=33 MT=102----- 5728 1451 88 Uncertainties in the evaluated excitation function for the 5728 1451 89 reaction La-139(n,g)La-140 are given in the three independent 5728 1451 90 matrixes. 5728 1451 91 In the energy range 1.000E-05 - 200 eV uncertainties are pre- 5728 1451 92 sented in the form of relative covariance matrix for the 16-neut- 5728 1451 93 ron energy groups (LB=5). Eigenvalues of this 6-th digits relat- 5728 1451 94 ive covariance matrix are the following: 5728 1451 95 5728 1451 96 1.29984E-04 1.80010E-04 2.41627E-04 2.69103E-04 5728 1451 97 3.03943E-04 3.53897E-04 4.83948E-04 5.72815E-04 5728 1451 98 6.17194E-04 6.90136E-04 9.46937E-04 1.01148E-03 5728 1451 99 1.61882E-03 1.76694E-03 1.98803E-03 2.89526E-02 5728 1451 100 5728 1451 101 In the energy range 0.2 - 20 keV uncertainties are given in 5728 1451 102 the form of diagonal matrix of uncertainties for 7-th neutron 5728 1451 103 energy intervals (LB=1) 5728 1451 104 In the energy range 20 keV - 20 MeV uncertainties are presen- 5728 1451 105 ted in the form of relative covariance matrix for the 23-neutron 5728 1451 106 energy groups (LB=5). Eigenvalues of this 6-th digits relative 5728 1451 107 covariance matrix are the following: 5728 1451 108 5728 1451 109 1.96223E-10 5.86297E-08 3.57093E-07 1.00052E-05 5728 1451 110 3.95061E-05 1.50141E-04 2.46548E-04 3.34259E-04 5728 1451 111 4.59370E-04 6.23219E-04 7.85798E-04 8.75929E-04 5728 1451 112 1.31621E-03 1.62913E-03 1.81439E-03 1.87643E-03 5728 1451 113 2.55332E-03 2.70062E-03 3.40954E-03 4.32177E-03 5728 1451 114 1.13231E-02 3.04915E-02 1.18873E-01 5728 1451 115 5728 1451 116 References : 5728 1451 117 1. S.F.Mughabghab et al. Neutron Cross Sections ,Vol.1,Part A, 5728 1451 118 New York, Academic Press,1981 5728 1451 119 2. S.I.Sukhoruchkin et al. Landolt Bornstein New Series, v.I/16B,5728 1451 120 ed. H.Schopper, Springer, 1998, pp.w19-w23 5728 1451 121 3. V.A.Konks, Ju.P.Popov, F.L.Shapiro Zurnal Eksperim. i Teor. 5728 1451 122 Fiziki, (USSR), v.46,(1), p.80, 1963 5728 1451 123 4. D.C.Stupegia et al. J. Nucl. Energ., v.22, p.267, May 1968 5728 1451 124 5. B.J.Allen, J.W.Boldeman, R.L.Macklin Nucl. Sci. Eng., v.82, 5728 1451 125 n.2, p.230, October 1982 5728 1451 126 6. G.N.Manturov et al. Voprosy Atomnoy Nauki i Tekhniki, Ser.: 5728 1451 127 Jadernye Konstanty, v.1, p.50, 1983 5728 1451 128 7. M.Divadeenam et al. Diss. Abstr. B , 1968, v.28, p.3834, 1968 5728 1451 129 8. E.Islam et al. Nucl. Phys. A, v.209, p.189, 1973 5728 1451 130 9. K.Nishimura et al. Report JAERI-M-6883, January 1977 5728 1451 131 10. R.L.Macklin et al. Phys. Rev., v.107, p.504, 1957 5728 1451 132 11. J.L.Perkin et al. Proc. Phys. Soc., v.72, p.505, 1958 5728 1451 133 12. W.S.Lyon, R.L.Macklin Phys. Rev., v.114, p.1619, 1959 5728 1451 134 13. A.E.Johnsrud et al. Phys. Rev., v.116, p.927, 1959 5728 1451 135 14. R.G.Wille, R.W.Fink Phys. Rev., v.118, p.242, 1960 5728 1451 136 15. J.H.Gibbons et al. Phys. Rev., v.122, p.182, 1961 5728 1451 137 16. J.Csikai et al. J. Atomki Kozlemenyek, v.8, p.79, June 1966 5728 1451 138 J.Csikai et al. Nucl. Phys. A, v.95, p.229, March 1967 5728 1451 139 17. A.K.Chaubey, M.L.Sehgal Phys. Rev., v.152, p.1055, Dec. 1966 5728 1451 140 18. G.Peto et al. J. Nucl. Energ., v.21, p.797, October 1967 5728 1451 141 19. F.Rigaud et al. Nucl. Phys., v.A176, p.545, December 1971 5728 1451 142 20. G.G.Zaikin et al. Ukrainskij Fizichnij Zhurnal, v.16(7), 5728 1451 143 p.1205, July 1971 5728 1451 144 21. F.Rigaud et al. Nucl. Sci. Eng., v.55, p.17, September 1974 5728 1451 145 22. O.Schwerer et al. Nucl. Phys. A, v.264, p.105, June 1976 5728 1451 146 23. A.R.Del.Musgrove, B.J.Allen, R.L.Macklin Proc. of Int. Conf. 5728 1451 147 on Neutron Physics and Nuclear Data for Reactors and Other 5728 1451 148 Applied Purposes, AERE Harwell, 25-29 September 1978, p.449 5728 1451 149 24. W.P.Poenitz Progress Report ANL-83-4, p.239, April 1982 5728 1451 150 25. J.Voignier et al. Nucl. Sci. Eng., v.93, p.43, 1986 5728 1451 151 26. H.Beer Progress Report NEANDC(E)-272U,(5), p.8, June 1986 5728 1451 152 27. Y.N.Trofimov Voprosy Atomnoy Nauki i Tekhniki, Serija: 5728 1451 153 Jadernye Konstanty, v.4, p.10, 1987 5728 1451 154 28. Y.N.Trofimov Proc. of the 1-st International Conf. on Neutron 5728 1451 155 Physics, Kiev, 14-18 September 1987, v.3, p.331, Moscow 1988 5728 1451 156 29. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 5728 1451 157 Model Code for Calculation of Cross Section and Emission 5728 1451 158 Spectra. Report LA-6947, Los Alamos, 1977 5728 1451 159 30. E.L.Trykov, G.Ya.Tertychnyi Private communication, IPPE, 5728 1451 160 Obninsk, May 1999 5728 1451 161 31. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 5728 1451 162 st's Meeting on Evaluation and Processing of Covariance Data, 5728 1451 163 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 5728 1451 164 32. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 5728 1451 165 33. S.F.Mughabghab et al. Termal Neutron Capture Cross Sections 5728 1451 166 Resonance Integrals and G-factors, Report INDC(NDS)-440, 5728 1451 167 IAEA, Vienna, February 2003 5728 1451 168 34. A.V.Ignatyuk et al. Landolt Bornstein New Series, v.I/16A, 5728 1451 169 Part 1, ed. H.Schopper, Springer, 1998, p.8-18 5728 1451 170 ***************************************************************** 5728 1451 171 The total and elastic cross-section files gave double counting 5728 1451 172 of the cross-sections in the resonance region, these two files 5728 1451 173 were modified for this library. The total and elastic cross- 5728 1451 174 section files were taken from ENDF/B-VI Release 1. 5728 1451 175 ***************************************************************** 5728 1451 176 ***************** Program LINEAR (VERSION 2002-1) ***************5728 1451 177 For All Data Greater than 1.0000E-10 barns in Absolute Value 5728 1451 178 Data Linearized to Within an Accuracy of .100000000 per-cent 5728 1451 179 ***************** Program RECENT (VERSION 2002-1) ***************5728 1451 180 for All Data Greater than 1.0000E-10 barns in Absolute Value 5728 1451 181 Data Linearized to within an Accuracy of .100000000 per-cent 5728 1451 182 ***************** Program SIGMA1 (VERSION 2002-1) ***************5728 1451 183 Data Doppler Broadened to 300.000000 Kelvin 5728 1451 184 for All Data Greater than 1.0000E-10 barns in Absolute Value 5728 1451 185 Data Linearized to Within an Accuracy pf .100000000 per-cent 5728 1451 186 ***************** Program FIXUP (Version 2002-1) ****************5728 1451 187 Corrected ZA/AWR in All Sections-----------------------------Yes 5728 1451 188 Corrected Thresholds-----------------------------------------Yes 5728 1451 189 Extended Cross Sections to 20 MeV----------------------------No 5728 1451 190 Allow Cross Section Deletion---------------------------------No 5728 1451 191 Allow Cross Section Reconstruction---------------------------No 5728 1451 192 Make All Cross Sections Non-Negative-------------------------Yes 5728 1451 193 Delete Energies Not in Ascending Order-----------------------Yes 5728 1451 194 Deleted Duplicate Points-------------------------------------Yes 5728 1451 195 Check for Ascending MAT/MF/MT Order--------------------------Yes 5728 1451 196 Check for Legal MF/MT Numbers--------------------------------Yes 5728 1451 197 Allow Creation of Missing Sections---------------------------No 5728 1451 198 Allow Insertion of Energy Points-----------------------------No 5728 1451 199 Create Uniform Energy Grid-----------------------------------No 5728 1451 200 Delete Section if Cross Section =0 at All Energies-----------Yes 5728 1451 201 ***************** Program GROUPIE (VERSION 2002-1) **************5728 1451 202 Unshielded Group Averages Using 640 Groups 5728 1451 203 Weighting Spectrum: Flat (Constant) Spectrum 5728 1451 204 1 451 208 15728 1451 205 2 151 4 15728 1451 206 3 102 217 15728 1451 207 33 102 93 15728 1451 208 5728 1 099999 5728 0 0 0 5.71390E+4 1.37713E+2 0 0 1 05728 2151 1 5.71390E+4 1.00000E+0 0 0 1 05728 2151 2 1.00000E-5 2.00000E+4 0 0 0 05728 2151 3 3.50000E+0 6.24850E-1 0 0 0 05728 2151 4 5728 2 099999 5728 0 0 0 5.71390E+4 1.37713E+2 0 0 0 05728 3102 1 5.16110E+6 5.16110E+6 0 0 1 6415728 3102 2 641 1 5728 3102 3 .000100000 140.916130 .000105000 137.631355 .000110000 134.5107985728 3102 4 .000115000 131.635331 .000120000 128.273838 .000127500 124.5443475728 3102 5 .000135000 121.137008 .000142500 118.005771 .000150000 114.6169365728 3102 6 .000160000 111.086689 .000170000 107.874636 .000180000 104.9262495728 3102 7 .000190000 102.183048 .000200000 99.6784848 .000210000 97.32212995728 3102 8 .000220000 95.1334787 .000230000 93.0999779 .000240000 90.71308485728 3102 9 .000255000 88.0902320 .000270000 86.0663388 .000280000 83.81339225728 3102 10 .000300000 81.0658839 .000320000 78.5781377 .000340000 76.29978725728 3102 11 .000360000 74.2013117 .000380000 72.2857027 .000400000 70.28859095728 3102 12 .000425000 68.2449527 .000450000 66.3776940 .000475000 64.66584655728 3102 13 .000500000 63.0560178 .000525000 61.5847291 .000550000 60.19486285728 3102 14 .000575000 58.9022361 .000600000 57.5773841 .000630000 56.21446155728 3102 15 .000660000 54.9673858 .000690000 53.7710963 .000720000 52.49819395728 3102 16 .000760000 51.1333567 .000800000 49.8661477 .000840000 48.70483885728 3102 17 .000880000 47.6010855 .000920000 46.5892455 .000960000 45.62022485728 3102 18 .001000000 44.6158143 .001050000 43.5673071 .001100000 42.58438595728 3102 19 .001150000 41.6787045 .001200000 40.6103452 .001275000 39.43618005728 3102 20 .001350000 38.3627031 .001425000 37.3608905 .001500000 36.29659065728 3102 21 .001600000 35.1837725 .001700000 34.1678926 .001800000 33.22832805728 3102 22 .001900000 32.3699820 .002000000 31.5741079 .002100000 30.82725945728 3102 23 .002200000 30.1432970 .002300000 29.4892640 .002400000 28.74300675728 3102 24 .002550000 27.9137684 .002700000 27.2651423 .002800000 26.56036255728 3102 25 .003000000 25.6927903 .003200000 24.9033544 .003400000 24.18153895728 3102 26 .003600000 23.5233241 .003800000 22.9156450 .004000000 22.28233845728 3102 27 .004250000 21.6394275 .004500000 21.0525874 .004750000 20.50405645728 3102 28 .005000000 20.0005519 .005250000 19.5336671 .005500000 19.09221685728 3102 29 .005750000 18.6875024 .006000000 18.2624172 .006300000 17.83749075728 3102 30 .006600000 17.4354503 .006900000 17.0627040 .007200000 16.65762145728 3102 31 .007600000 16.2246636 .008000000 15.8289941 .008400000 15.45792825728 3102 32 .008800000 15.1129611 .009200000 14.7923717 .009600000 14.48729375728 3102 33 .010000000 14.1711393 .010500000 13.8366769 .011000000 13.53026115728 3102 34 .011500000 13.2378410 .012000000 12.9026382 .012750000 12.53202545728 3102 35 .013500000 12.1890031 .014250000 11.8743787 .015000000 11.53881425728 3102 36 .016000001 11.1865003 .017000001 10.8633443 .017999999 10.56820955728 3102 37 .018999999 10.2967876 .020000000 10.0420336 .021000000 9.808730435728 3102 38 .022000000 9.58877507 .023000000 9.38366445 .024000000 9.142476865728 3102 39 .025500000 8.88195969 .027000001 8.67540086 .028000001 8.450706555728 3102 40 .029999999 8.17372671 .032000002 7.92099010 .034000002 7.689985845728 3102 41 .035999998 7.47901504 .037999999 7.28440094 .039999999 7.081330265728 3102 42 .042500000 6.87522787 .045000002 6.68700537 .047499999 6.511801765728 3102 43 .050000001 6.35088582 .052499998 6.20203607 .055000000 6.061114085728 3102 44 .057500001 5.93184964 .059999999 5.79580688 .063000001 5.659533175728 3102 45 .066000000 5.53106507 .068999998 5.41136298 .071999997 5.281554185728 3102 46 .075999998 5.14250084 .079999998 5.01545054 .083999999 4.896032725728 3102 47 .088000000 4.78499161 .092000000 4.68226134 .096000001 4.584216205728 3102 48 .100000001 4.48324276 .104999997 4.37641057 .109999999 4.278466245728 3102 49 .115000002 4.18477460 .119999997 4.07719395 .127499998 3.958265675728 3102 50 .135000005 3.84797071 .142499998 3.74673540 .150000006 3.638811005728 3102 51 .159999996 3.52534063 .170000002 3.42151836 .180000007 3.326642955728 3102 52 .189999998 3.23955461 .200000003 3.15757224 .209999993 3.082848935728 3102 53 .219999999 3.01500584 .230000004 2.94815628 .239999995 2.871692795728 3102 54 .254999995 2.78666076 .270000011 2.72163373 .280000001 2.649409865728 3102 55 .300000012 2.56098805 .319999993 2.48088377 .340000004 2.408141305728 3102 56 .360000014 2.34167369 .379999995 2.27990170 .400000006 2.216861425728 3102 57 .425000012 2.15320686 .449999988 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2.441640-3 9.596400-3 2.815990-3 9.693480-4 1.103800-3 1.023690-3572833102 73 1.074000-3 1.221770-3 1.341960-3 1.284610-3 1.173230-3 1.279090-3572833102 74 1.172100-3 1.398410-3 1.645660-3 2.135440-3 5.787750-3 1.657310-3572833102 75 2.205260-3 2.617220-3 2.669340-3 2.556840-3 2.464170-3 2.529390-3572833102 76 2.598890-3 2.555700-3 2.531740-3 2.824340-3 3.602280-3 4.223850-3572833102 77 3.633110-3 4.267780-3 4.641970-3 4.800960-3 4.840430-3 4.849650-3572833102 78 4.869290-3 4.846410-3 4.898520-3 4.754970-3 5.395380-3 6.744660-3572833102 79 8.112480-3 5.966690-3 6.739140-3 6.708790-3 6.302550-3 5.970150-3572833102 80 6.109290-3 6.347980-3 6.211780-3 6.184360-3 6.864580-3 8.790700-3572833102 81 1.027570-2 8.063440-3 8.221160-3 7.668950-3 7.008490-3 6.952500-3572833102 82 7.356010-3 7.186020-3 7.197940-3 7.924890-3 1.020480-2 1.189920-2572833102 83 8.717380-3 8.454870-3 7.820360-3 7.406320-3 7.592980-3 7.632630-3572833102 84 7.502130-3 8.380310-3 1.061400-2 1.263480-2 8.701350-3 8.438740-3572833102 85 7.794530-3 7.432500-3 7.718140-3 7.386790-3 8.453560-3 1.047490-2572833102 86 1.273760-2 8.742840-3 8.359810-3 7.410060-3 7.578710-3 7.242570-3572833102 87 8.353630-3 1.032520-2 1.250930-2 8.829350-3 8.074660-3 7.385510-3572833102 88 7.413820-3 8.264620-3 1.053090-2 1.236150-2 8.774640-3 7.717270-3572833102 89 7.483530-3 8.327420-3 1.057710-2 1.255790-2 8.949390-3 7.595650-3572833102 90 8.383470-3 1.059340-2 1.256530-2 8.513730-3 8.420100-3 1.022820-2572833102 91 1.243340-2 1.077640-2 1.217860-2 1.364410-2 1.682810-2 1.804110-2572833102 92 2.429660-2 572833102 93 572833 099999 5728 0 0 0 0 0 0 0 5.91410E+4 1.39697E+2 0 0 34 105925 1451 1 0.0 0.0 0 0 0 65925 1451 2 1.00000E+0 2.00000E+7 0 0 10 20025925 1451 3 3.00000E+2 0.0 1 0 122 45925 1451 4 59-Pr-141 FEI EVAL-Jan97 K.I.Zolotarev et al. 5925 1451 5 DIST-Feb2004 5925 1451 6 ----IRDF-2002 MATERIAL 5925 5925 1451 7 -----INCIDENT NEUTRON DATA 5925 1451 8 ------ENDF-6 FORMAT 5925 1451 9 ***************************************************************** 5925 1451 10 59-PR-141 FEI EVAL-Jan97 K.I.Zolotarev et al. 5925 1451 11 DIST-Jan02 20020123 5925 1451 12 ----BROND-2 MATERIAL 5925 5925 1451 13 -----INCIDENT NEUTRON DATA 5925 1451 14 ------ENDF-6 FORMAT 5925 1451 15 ------Russian Reactor Dosimetry File RRDF-2002 5925 1451 16 ***************************************************************** 5925 1451 17 Authors of evaluation: K.Zolotarev, V.Manokhin, A.Pashchenko 5925 1451 18 ***************************************************************** 5925 1451 19 MF= 3 5925 1451 20 MT= 16 - Pr-141(n,2n)Pr-140 reaction 5925 1451 21 -------------------------------------- 5925 1451 22 Excitation function for the Pr-141(n,2n)Pr-140 reaction in 5925 1451 23 the energy region from threshold to 20 MeV was evaluated by means 5925 1451 24 of statistical analysis of experimental cross section data [1-19] 5925 1451 25 and data from systematics of (n,2n) excitation function [20] in 5925 1451 26 the energy regions 9.5 - 12.0 MeV and 18.0 - 20.0 MeV. 5925 1451 27 All experimental data were renormalized to the new standards 5925 1451 28 for monitor reactions cross sections and decay data. 5925 1451 29 The final procedure of evaluation has been carried out within 5925 1451 30 the framework of generalized least squares method. Rational func- 5925 1451 31 tion was used as model function [21]. Calculations was performed 5925 1451 32 by means of Pade-2 code [22]. 5925 1451 33 U-235 thermal fission [23] and Cf-252 spontaneous fission 5925 1451 34 neutron spectra [24] averaged cross-sections calculated from the 5925 1451 35 evaluated Pr141(n,2n)Pr140 excitation function are the following: 5925 1451 36 5925 1451 37 -------------------------------------------- 5925 1451 38 TYPE OF SPECTRUM I CS>, mb (calc.) 5925 1451 39 --------------------------I----------------- 5925 1451 40 U-235 neutron fission I 1.0922 5925 1451 41 CF-252 spontan. fission I 1.9843 5925 1451 42 5925 1451 43 MF=33 5925 1451 44 MT= 16 -(n,2n) cross section cov. matrix 5925 1451 45 ---------------------------------------- 5925 1451 46 Uncertainties in the evaluated excitation function for the re-5925 1451 47 action Pr-141(n,2n)Pr-140 are given in the form of relative cova- 5925 1451 48 riance matrix for the 18-neutron energy groups (LB=5). Covariance 5925 1451 49 matrix of uncertainties was calculated simultaneously with 5925 1451 50 recommended cross section data by means of PADE-2 code. 5925 1451 51 Eigenvalues of the 6-th digits relative covariance matrix 5925 1451 52 given in the 33-file are the following: 5925 1451 53 5925 1451 54 6.47369E-06 6.50089E-06 6.56924E-06 6.70314E-06 5925 1451 55 6.90792E-06 7.20594E-06 7.64163E-06 8.30022E-06 5925 1451 56 9.35563E-06 1.12079E-05 1.49213E-05 2.36154E-05 5925 1451 57 4.48092E-05 3.28699E-04 6.81059E-03 1.70941E-02 5925 1451 58 2.55798E-02 1.53270E-01 5925 1451 59 5925 1451 60 References : 5925 1451 61 1. E.B.Paul, R.L.Clarke Canadian J. Phys., v.31,p.267, 1953 5925 1451 62 2. J.M.Ferguson, W.E.Thompson Phys. Rev., v.118, p.228, 1960 5925 1451 63 3. R.G.Wille, R.W.Fink Phys. Rev., v.118, p.242, 1960 5925 1451 64 4. L.A.Rayburn Phys. Rev., v.122, p.168, 1961 5925 1451 65 5. C.S.Khurana, H.S.Hans Nucl. Phys., v.28, p.560, 1961 5925 1451 66 6. D.R.Koehler, W.L.Alford Report NP-11667, 1962 5925 1451 67 7. M.Cevolani, S.Petralia Nucl. Sci. Eng., v.26, p.1328, 1962 5925 1451 68 8. L.A.Rayburn Bull. American Phys. Soc., v.8, p.60, Jan. 1963 5925 1451 69 9. M.P.Menon, M.Y.Cuypers Phys. Rev., v.156, p.1340, 1967 5925 1451 70 10. P.Cuzzocrea e.a. Nuovo Cimento, Sec.B, v.52, n.2, p.476, 1967 5925 1451 71 11. M.Bormann e.a. Nucl. Phys., Sec.A, v.115, p.309, July 1968 5925 1451 72 12. G.Peto e.a. Acta Physica Hung., v.24, p.93, April 1968 5925 1451 73 13. A.Chatterjee e.a. Proc. of 12th Nucl. Phys. and Solid State 5925 1451 74 Phys. Sympos., Roorkee, India, 28-31 December 1969, v.2,p.117 5925 1451 75 14. A.Bari Dissertation Abstracts, sec.B, v.32, p.5091, 1972 5925 1451 76 15. J.Araminowicz, J.Dresler Prog. Rep. INR-1464, p.14, May 1973 5925 1451 77 16. R.A.Sigg, P.K.Kuroda Inorg. Nucl. Chem., v.37, p.631, 1975 5925 1451 78 17. M.Valkonen Report JU-RR-1/1976, Jyvaeskylae University, 1976 5925 1451 79 18. S.Murahira et al. Report INDC(JPN)-175/U, p.171, Nov. 1995 5925 1451 80 19. Y.Kasugai, Y.Ikeda, Y.Uno Proc. of Int. Conf. on Nuclear Data 5925 1451 81 for Science and Technology, Trieste, Italy, 19-24 May 1997, 5925 1451 82 v.1, p.635 5925 1451 83 20. V.N.Manokhin Report INDC(CCP)-397, IAEA, Vienna, 1997 5925 1451 84 21. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 5925 1451 85 st's Meeting on Evaluation and Processing of Covariance Data, 5925 1451 86 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 5925 1451 87 22. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 5925 1451 88 23. L.W.Weston et al. Evaluated Neutron Data for U-235, ENDF/B-VI 5925 1451 89 Library, MAT=9228, MF=5, MT=18, eval. April 1989 5925 1451 90 24. W.Mannhart IAEA-TECDOC-410, p.158, Vienna, 1987 5925 1451 91 ******************************************************************5925 1451 92 The Q values and threshold energies were updated prior to pro- 5925 1451 93 cessing through the codes to comply with the values obtained 5925 1451 94 using the NNDC calculation program which is based on the 1995 5925 1451 95 Update to the Atomic mass Evaluation. 5925 1451 96 5925 1451 97 File 2 added to the pointwise file containing only the effective 5925 1451 98 scattering radius with no resonance parameters given. 5925 1451 99 Taken from ENDF/B-VI 5925 1451 100 ******************************************************************5925 1451 101 ***************** Program LINEAR (VERSION 2002-1) ***************5925 1451 102 For All Data Greater than 1.0000E-10 barns in Absolute Value 5925 1451 103 Data Linearized to Within an Accuracy of .100000000 per-cent 5925 1451 104 ***************** Program SIGMA1 (VERSION 2002-1) ***************5925 1451 105 Data Doppler Broadened to 300.000000 Kelvin 5925 1451 106 for All Data Greater than 1.0000E-10 barns in Absolute Value 5925 1451 107 Data Linearized to Within an Accuracy pf .100000000 per-cent 5925 1451 108 ***************** Program FIXUP (Version 2002-1) ****************5925 1451 109 Corrected ZA/AWR in All Sections-----------------------------Yes 5925 1451 110 Corrected Thresholds-----------------------------------------Yes 5925 1451 111 Extended Cross Sections to 20 MeV----------------------------No 5925 1451 112 Allow Cross Section Deletion---------------------------------No 5925 1451 113 Allow Cross Section Reconstruction---------------------------No 5925 1451 114 Make All Cross Sections Non-Negative-------------------------Yes 5925 1451 115 Delete Energies Not in Ascending Order-----------------------Yes 5925 1451 116 Deleted Duplicate Points-------------------------------------Yes 5925 1451 117 Check for Ascending MAT/MF/MT Order--------------------------Yes 5925 1451 118 Check for Legal MF/MT Numbers--------------------------------Yes 5925 1451 119 Allow Creation of Missing Sections---------------------------No 5925 1451 120 Allow Insertion of Energy Points-----------------------------No 5925 1451 121 Create Uniform Energy Grid-----------------------------------No 5925 1451 122 Delete Section if Cross Section =0 at All Energies-----------Yes 5925 1451 123 ***************** Program GROUPIE (VERSION 2002-1) **************5925 1451 124 Unshielded Group Averages Using 640 Groups 5925 1451 125 Weighting Spectrum: Flat (Constant) Spectrum 5925 1451 126 1 451 130 15925 1451 127 2 151 4 15925 1451 128 3 16 39 15925 1451 129 33 16 38 15925 1451 130 5925 1 099999 5925 0 0 0 5.91410E+4 1.39697E+2 0 0 1 05925 2151 1 5.914100+4 1.000000+0 0 0 1 05925 2151 2 1.000000-5 1.004920+4 0 0 0 05925 2151 3 2.500000+0 4.930000-1 0 0 0 05925 2151 4 5925 2 099999 5925 0 0 0 5.91410E+4 1.39697E+2 0 0 0 05925 3 16 1 -9.39700E+6-9.39700E+6 0 0 1 1075925 3 16 2 107 1 5925 3 16 3 9400000.00 .000102039 9500000.00 .004115304 9600000.00 .0164385955925 3 16 4 9700000.00 .038856950 9800000.00 .070499450 9900000.00 .1102883505925 3 16 5 10000000.0 .157013000 10100000.0 .209402000 10200000.0 .2661885005925 3 16 6 10300000.0 .326164000 10400000.0 .388315750 10500000.0 .4513412505925 3 16 7 10600000.0 .514366750 10700000.0 .577392250 10800000.0 .6395045005925 3 16 8 10900000.0 .699850500 11000000.0 .758343500 11100000.0 .8147340005925 3 16 9 11200000.0 .868854500 11300000.0 .920604500 11400000.0 .9699370005925 3 16 10 11500000.0 1.01684800 11600000.0 1.06136500 11700000.0 1.102987505925 3 16 11 11800000.0 1.14290250 11900000.0 1.18068500 12000000.0 1.216335005925 3 16 12 12100000.0 1.25004500 12200000.0 1.28181500 12300000.0 1.311847505925 3 16 13 12400000.0 1.34014250 12500000.0 1.36689000 12600000.0 1.392090005925 3 16 14 12700000.0 1.41561500 12800000.0 1.43746500 12900000.0 1.459315005925 3 16 15 13000000.0 1.47946167 13100000.0 1.49790500 13200000.0 1.516348335925 3 16 16 13300000.0 1.53338000 13400000.0 1.54900000 13500000.0 1.564620005925 3 16 17 13600000.0 1.57889750 13700000.0 1.59183250 13800000.0 1.604767505925 3 16 18 13900000.0 1.61770250 14000000.0 1.62927600 14100000.0 1.639488005925 3 16 19 14200000.0 1.64970000 14300000.0 1.65991200 14400000.0 1.670124005925 3 16 20 14500000.0 1.67916600 14600000.0 1.68703800 14700000.0 1.694910005925 3 16 21 14800000.0 1.70278200 14900000.0 1.71065400 15000000.0 1.717540835925 3 16 22 15100000.0 1.72344250 15200000.0 1.72934417 15300000.0 1.735245835925 3 16 23 15400000.0 1.74114750 15500000.0 1.74704917 15600000.0 1.752039295925 3 16 24 15700000.0 1.75611786 15800000.0 1.76019643 15900000.0 1.764275005925 3 16 25 16000000.0 1.76835357 16100000.0 1.77243214 16200000.0 1.776510715925 3 16 26 16300000.0 1.77971063 16400000.0 1.78203188 16500000.0 1.784353135925 3 16 27 16600000.0 1.78667438 16700000.0 1.78899563 16800000.0 1.791316885925 3 16 28 16900000.0 1.79363813 17000000.0 1.79595938 17100000.0 1.797471675925 3 16 29 17200000.0 1.79817500 17300000.0 1.79887833 17400000.0 1.799581675925 3 16 30 17500000.0 1.80028500 17600000.0 1.80098833 17700000.0 1.800834175925 3 16 31 17800000.0 1.79982250 17900000.0 1.79881083 18000000.0 1.797799175925 3 16 32 18100000.0 1.79678750 18200000.0 1.79577583 18300000.0 1.793587005925 3 16 33 18400000.0 1.79022100 18500000.0 1.78685500 18600000.0 1.783489005925 3 16 34 18700000.0 1.78012300 18800000.0 1.77536333 18900000.0 1.769210005925 3 16 35 19000000.0 1.76305667 19100000.0 1.75515667 19200000.0 1.745510005925 3 16 36 19300000.0 1.73586333 19400000.0 1.72389250 19500000.0 1.709597505925 3 16 37 19600000.0 1.69321500 19700000.0 1.67284000 19800000.0 1.648080005925 3 16 38 19900000.0 1.61751500 20000000.0 0.0 5925 3 16 39 5925 3 099999 5925 0 0 0 5.91410E+4 1.39697E+2 0 0 0 1592533 16 1 0.000000+0 0.000000+0 0 16 0 1592533 16 2 0.000000+0 0.000000+0 1 5 210 20592533 16 3 1.000000-5 9.400000+6 1.050000+7 1.100000+7 1.150000+7 1.200000+7592533 16 4 1.250000+7 1.300000+7 1.350000+7 1.400000+7 1.450000+7 1.500000+7592533 16 5 1.550000+7 1.600000+7 1.650000+7 1.700000+7 1.750000+7 1.800000+7592533 16 6 1.850000+7 1.900000+7 0.000000+0 0.000000+0 0.000000+0 0.000000+0592533 16 7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0592533 16 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0592533 16 9 0.000000+0 0.000000+0 0.000000+0 7.629710-2 5.270640-2 3.376320-2592533 16 10 2.074520-2 1.252350-2 7.563810-3 4.706180-3 3.188710-3 2.533760-3592533 16 11 2.446920-3 2.749080-3 3.334790-3 4.148410-3 5.171510-3 6.419360-3592533 16 12 7.945970-3 9.861990-3 1.237550-2 4.016490-2 2.809550-2 1.900490-2592533 16 13 1.265780-2 8.366950-3 5.519320-3 3.674800-3 2.539500-3 1.923500-3592533 16 14 1.707930-3 1.823370-3 2.237300-3 2.948350-3 3.986210-3 5.417190-3592533 16 15 7.354640-3 9.962330-3 2.145550-2 1.571000-2 1.127530-2 7.970050-3592533 16 16 5.552600-3 3.813640-3 2.597590-3 1.796650-3 1.340690-3 1.189230-3592533 16 17 1.326580-3 1.760550-3 2.524460-3 3.682770-3 5.338130-3 7.621620-3592533 16 18 1.236030-2 9.413210-3 7.029290-3 5.152190-3 3.707030-3 2.623120-3592533 16 19 1.845220-3 1.335170-3 1.071090-3 1.046710-3 1.272260-3 1.777090-3592533 16 20 2.614640-3 3.867430-3 5.635080-3 7.582390-3 5.943050-3 4.583860-3592533 16 21 3.484360-3 2.620050-3 1.966010-3 1.503080-3 1.220170-3 1.115420-3592533 16 22 1.197700-3 1.489140-3 2.029030-3 2.877480-3 4.106360-3 4.916750-3592533 16 23 3.993780-3 3.224900-3 2.597250-3 2.101980-3 1.730500-3 1.477540-3592533 16 24 1.342690-3 1.331920-3 1.459470-3 1.750620-3 2.244570-3 2.989940-3592533 16 25 3.456170-3 2.967610-3 2.561130-3 2.226210-3 1.959650-3 1.758980-3592533 16 26 1.623750-3 1.556590-3 1.564460-3 1.660280-3 1.865240-3 2.208980-3592533 16 27 2.737410-3 2.515350-3 2.327170-3 2.163680-3 2.023420-3 1.905210-3592533 16 28 1.808580-3 1.734270-3 1.685040-3 1.667360-3 1.695630-3 2.469300-3592533 16 29 2.400280-3 2.330770-3 2.251710-3 2.160910-3 2.055290-3 1.930710-3592533 16 30 1.782030-3 1.604310-3 1.399900-3 2.450900-3 2.455830-3 2.435060-3592533 16 31 2.378170-3 2.280110-3 2.133270-3 1.926860-3 1.647670-3 1.288780-3592533 16 32 2.543100-3 2.568940-3 2.550460-3 2.474410-3 2.331110-3 2.106380-3592533 16 33 1.781790-3 1.343610-3 2.656650-3 2.673470-3 2.633140-3 2.518610-3592533 16 34 2.314590-3 2.000500-3 1.558140-3 2.749450-3 2.752370-3 2.692880-3592533 16 35 2.550410-3 2.305410-3 1.937640-3 2.834100-3 2.852250-3 2.816520-3592533 16 36 2.705480-3 2.499580-3 3.001600-3 3.118840-3 3.217400-3 3.275390-3592533 16 37 3.472370-3 3.866030-3 4.312850-3 4.687830-3 5.673930-3 7.407790-3592533 16 38 592533 099999 5925 0 0 0 0 0 0 0 6.40000E+4 1.55901E+2 -1 0 34 106400 1451 1 0.0 0.0 0 0 0 66400 1451 2 1.00000E+0 2.00000E+7 0 0 10 20026400 1451 3 3.00000E+2 0.0 1 0 59 36400 1451 4 64-Gd- 0 NDS IAEA-JAN04 6400 1451 5 DIST-Feb2004 6400 1451 6 ----IRDF-2002 6400 1451 7 -----INCIDENT NEUTRON DATA 6400 1451 8 ------ENDF-6 FORMAT 6400 1451 9 *****************************************************************6400 1451 10 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************6400 1451 11 DOSIMETRY Assembled at NDS from the ENDF/B-VI evaluations for 6400 1451 12 64-Gd-152,154,155,156,157,158 and 160. The files 6400 1451 13 were processed through the 2002 pre-processing codes 6400 1451 14 LINEAR,RECENT,SIGMA1 and FIXUP to produce pointwise 6400 1451 15 files prior to input to MIXER. 6400 1451 16 6400 1451 17 MIXER INPUT FILE 6400 1451 18 ***************** Program MIXER (VERSION 2002-1) ****************6400 1451 19 GADOLINIUM,MET. DENSITY 7.901 G/CC. CONSTITUENTS: 160-21.86 6400 1451 20 158-24.84, 152-0.2, 154-2.18, 155-14.8,156-20.47, 157-15.65 6400 1451 21 ---------------------------------------- 6400 1451 22 Composition 6400 1451 23 ---------------------------------------- 6400 1451 24 Isotope MF MT Atom-Fract Grams/cc 6400 1451 25 ---------------------------------------- 6400 1451 26 64152 3 1 .001999808 .015264700 6400 1451 27 64154 3 1 .021799790 .168591500 6400 1451 28 64155 3 1 .147999470 1.15202110 6400 1451 29 64156 3 1 .204700250 1.60365810 6400 1451 30 64157 3 1 .156499785 1.23392290 6400 1451 31 64158 3 1 .248400703 1.97099140 6400 1451 32 64160 3 1 .218600194 1.75653450 6400 1451 33 ---------------------------------------- 6400 1451 34 64000 3 1 1.00000000 7.90098420 6400 1451 35 ---------------------------------------- 6400 1451 36 ***************** Program MIXER (VERSION 2002-1) ****************6400 1451 37 GADOLINIUM,MET. DENSITY 7.901 G/CC. CONSTITUENTS: 160-21.86 6400 1451 38 158-24.84, 152-0.2, 154-2.18, 155-14.8,156-20.47, 157-15.65 6400 1451 39 ---------------------------------------- 6400 1451 40 Composition 6400 1451 41 ---------------------------------------- 6400 1451 42 Isotope MF MT Atom-Fract Grams/cc 6400 1451 43 ---------------------------------------- 6400 1451 44 64152 3 102 .001999808 .015264700 6400 1451 45 64154 3 102 .021799790 .168591500 6400 1451 46 64155 3 102 .147999470 1.15202110 6400 1451 47 64156 3 102 .204700250 1.60365810 6400 1451 48 64157 3 102 .156499785 1.23392290 6400 1451 49 64158 3 102 .248400703 1.97099140 6400 1451 50 64160 3 102 .218600194 1.75653450 6400 1451 51 ---------------------------------------- 6400 1451 52 64000 3 102 1.00000000 7.90098420 6400 1451 53 ---------------------------------------- 6400 1451 54 *****************************************************************6400 1451 55 6400 1451 56 6400 1451 57 6400 1451 58 6400 1451 59 *****************************************************************6400 1451 60 ***************** Program GROUPIE (VERSION 2002-1) **************6400 1451 61 Unshielded Group Averages Using 640 Groups 6400 1451 62 Weighting Spectrum: Flat (Constant) Spectrum 6400 1451 63 1 451 66 06400 1451 64 3 1 217 06400 1451 65 3 102 217 06400 1451 66 6400 1 099999 6400 0 0 0 6.40000E+4 1.55901E+2 0 0 0 06400 3 1 1 0.0 0.0 0 0 1 6416400 3 1 2 641 1 6400 3 1 3 .000100000 589975.698 .000105000 576263.743 .000110000 563237.9996400 3 1 4 .000115000 551236.393 .000120000 537207.265 .000127500 521643.7756400 3 1 5 .000135000 507426.347 .000142500 494362.490 .000150000 480225.4426400 3 1 6 .000160000 465500.570 .000170000 452104.974 .000180000 439810.8086400 3 1 7 .000190000 428373.756 .000200000 417933.482 .000210000 408112.2456400 3 1 8 .000220000 398991.461 .000230000 390518.607 .000240000 380574.7966400 3 1 9 .000255000 369650.434 .000270000 361222.513 .000280000 351842.7476400 3 1 10 .000300000 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.003124347 20000000.0 0.0 6400 3102 217 6400 3 099999 6400 0 0 0 0 0 0 0 6.91690E+4 1.67483E+2 0 0 34 106925 1451 1 0.0 0.0 0 0 0 66925 1451 2 1.00000E+0 2.00000E+7 0 0 10 20026925 1451 3 3.00000E+2 0.0 1 0 88 46925 1451 4 69-Tm-169 SRI EVAL-JAN96 N.ODANO (SHIP RES. INST.) 6925 1451 5 DIST-Feb2004 6925 1451 6 ----IRDF-2002 MATERIAL 6925 6925 1451 7 -----INCIDENT NEUTRON DATA 6925 1451 8 ------ENDF-6 FORMAT 6925 1451 9 **************************************************************** 6925 1451 10 69-TM-169 SRI EVAL-JAN96 N.ODANO (SHIP RES. INST.) 6925 1451 11 DIST-JUL98 6925 1451 12 ----JENDL/D-99 MATERIAL 6925 6925 1451 13 **************************************************************** 6925 1451 14 HISTORY 6925 1451 15 96-01 EVALUATION FOR JENDL DOSIMETRY FILE VERSION 2 WAS MADE BY 6925 1451 16 N.ODANO (SHIP RES. INST.). 6925 1451 17 97-09 COMPILED TO JENDL DOSIMETRY FILE 99. 6925 1451 18 6925 1451 19 ==== POINT-WISE DATA FILE ==== 6925 1451 20 6925 1451 21 6925 1451 22 TM-169 (N,2N) TM-168 (HALF-LIFE = 93.1 D) 6925 1451 23 6925 1451 24 MF=1 GENERAL INFORMATION 6925 1451 25 MT=451 DESCRIPTIVE DATA AND DICTIONARY 6925 1451 26 6925 1451 27 MF=2 RESONANCE PARAMETERS 6925 1451 28 MT=151 PARAMETERS 6925 1451 29 ONLY SPIN AND SCATTERING RADIUS ARE GIVEN. 6925 1451 30 6925 1451 31 MF=3 NEUTRON CROSS SECTIONS 6925 1451 32 MT=16 (N,2N) CROSS SECTION 6925 1451 33 EXPERIMENTAL DATA/1-13/ IN NESTOR-2/14/ WERE TAKEN FOR 6925 1451 34 THE EVALUATION USING GMA CODE/15/. 6925 1451 35 6925 1451 36 MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS 6925 1451 37 MT=16 GENERATED USING THE GMA CODE. 6925 1451 38 6925 1451 39 REFERENCES 6925 1451 40 1) W.DILG+ : NUCL. PHYS. A, 118, 9 (1968). 6925 1451 41 2) D.S.MATHER+ : AWRE-O-72/,72 (1972). 6925 1451 42 3) D.R.NETHERWAY : NUCL. PHYS. A, 190, 635 (1972). 6925 1451 43 4) R.VOS+ : BULLETIN OF THE AMERICAN PHYSICAL SOCIETY, 18, 6925 1451 44 775(EB6) (1973). 6931 1451 6925 1451 45 5) S.M.QAIM : NUCL. PHYS. A, 224, 319 (1974). 6925 1451 46 6) B.P.BAYHURST+ : PHYS. REV. C, 12, 451 (1975). 6925 1451 47 7) L.R.VEESER+ : PHYS. REV. C, 16, 1792 (1977). 6925 1451 48 8) J.FREHAUT+ : PROC. SYMP. ON NEUTRON CROSS SECTIONS FROM 10-50 6925 1451 49 MEV, UPTON, USA, 12-14 MAY, 1980, P.399 (1980). 6925 1451 50 9) J.LAUREC+ : CEA-R-5109 (1981). 6925 1451 51 10) L.R.GREENWOOD : ASTM-STP-956, 743 (1987). 6925 1451 52 11) LU HANLIN+ : INDC(CRP)-16 (1989). 6925 1451 53 12) WANG XIUYUAN+ : A COLLECTION OF ACADEMIC EXCHANGE REPORT IN 6925 1451 54 THE FIELDS OD RADIOCHEMISTRY, CHEMICAL INDUSTRY AND ISOTOPE, 6925 1451 55 NUCLEAR INSTITUTE OF SICHUAN (1989). 6925 1451 56 13) C.KONNO+ : JAERI 1329 (1993). 6925 1451 57 14) T.NAKAGAWA : THE JAERI NUCLEAR DATA CENTER, UNPUBLISHED. 6925 1451 58 15) W.P.POENITZ : PROC. CONF. NUCLEAR DATA EVALUATION METHODS 6925 1451 59 AND PROCEDURES, BROOKHAVEN NATIONAL LAB. 1980, BNL-NCS- 6925 1451 60 51363, P.249 (1981). 6925 1451 61 **************************************************************** 6925 1451 62 The Q values and threshold energies were updated prior to pro- 6925 1451 63 cessing through the codes to comply with the values obtained 6925 1451 64 using the NNDC calculation program which is based on the 1995 6925 1451 65 Update to the Atomic mass Evaluation. 6925 1451 66 **************************************************************** 6925 1451 67 ***************** Program LINEAR (VERSION 2002-1) ***************6925 1451 68 For All Data Greater than 1.0000E-10 barns in Absolute Value 6925 1451 69 Data Linearized to Within an Accuracy of .100000000 per-cent 6925 1451 70 ***************** Program SIGMA1 (VERSION 2002-1) ***************6925 1451 71 Data Doppler Broadened to 300.000000 Kelvin 6925 1451 72 for All Data Greater than 1.0000E-10 barns in Absolute Value 6925 1451 73 Data Linearized to Within an Accuracy pf .100000000 per-cent 6925 1451 74 ***************** Program FIXUP (Version 2002-1) ****************6925 1451 75 Corrected ZA/AWR in All Sections-----------------------------Yes 6925 1451 76 Corrected Thresholds-----------------------------------------Yes 6925 1451 77 Extended Cross Sections to 20 MeV----------------------------No 6925 1451 78 Allow Cross Section Deletion---------------------------------No 6925 1451 79 Allow Cross Section Reconstruction---------------------------No 6925 1451 80 Make All Cross Sections Non-Negative-------------------------Yes 6925 1451 81 Delete Energies Not in Ascending Order-----------------------Yes 6925 1451 82 Deleted Duplicate Points-------------------------------------Yes 6925 1451 83 Check for Ascending MAT/MF/MT Order--------------------------Yes 6925 1451 84 Check for Legal MF/MT Numbers--------------------------------Yes 6925 1451 85 Allow Creation of Missing Sections---------------------------No 6925 1451 86 Allow Insertion of Energy Points-----------------------------No 6925 1451 87 Create Uniform Energy Grid-----------------------------------No 6925 1451 88 Delete Section if Cross Section =0 at All Energies-----------Yes 6925 1451 89 ***************** Program GROUPIE (VERSION 2002-1) **************6925 1451 90 Unshielded Group Averages Using 640 Groups 6925 1451 91 Weighting Spectrum: Flat (Constant) Spectrum 6925 1451 92 1 451 96 16925 1451 93 2 151 4 16925 1451 94 3 16 44 16925 1451 95 33 16 23 16925 1451 96 6925 1 099999 6925 0 0 0 6.91690E+4 1.67483E+2 0 0 1 06925 2151 1 6.91690E+4 1.00000E+0 0 0 1 06925 2151 2 0.0 0.0 0 0 0 06925 2151 3 5.00000E-1 7.70000E-1 0 0 0 06925 2151 4 6925 2 099999 6925 0 0 0 6.91690E+4 1.67483E+2 0 0 0 06925 3 16 1 -8.03340E+6-8.03340E+6 0 0 1 1216925 3 16 2 121 1 6925 3 16 3 8000000.00 .000162775 8100000.00 .006767183 8200000.00 .0167043106925 3 16 4 8300000.00 .026641437 8400000.00 .070615833 8500000.00 .1486275006925 3 16 5 8600000.00 .226639167 8700000.00 .304650833 8800000.00 .3826625006925 3 16 6 8900000.00 .460674167 9000000.00 .535816000 9100000.00 .6080880006925 3 16 7 9200000.00 .680360000 9300000.00 .752632000 9400000.00 .8249040006925 3 16 8 9500000.00 .897176000 9600000.00 .969448000 9700000.00 1.041720006925 3 16 9 9800000.00 1.11399200 9900000.00 1.18626400 10000000.0 1.235935006925 3 16 10 10100000.0 1.26300500 10200000.0 1.29007500 10300000.0 1.317145006925 3 16 11 10400000.0 1.34421500 10500000.0 1.37128500 10600000.0 1.398355006925 3 16 12 10700000.0 1.42542500 10800000.0 1.45249500 10900000.0 1.479565006925 3 16 13 11000000.0 1.50758000 11100000.0 1.53654000 11200000.0 1.565500006925 3 16 14 11300000.0 1.59446000 11400000.0 1.62342000 11500000.0 1.652380006925 3 16 15 11600000.0 1.68134000 11700000.0 1.71030000 11800000.0 1.739260006925 3 16 16 11900000.0 1.76822000 12000000.0 1.79406500 12100000.0 1.816795006925 3 16 17 12200000.0 1.83952500 12300000.0 1.86225500 12400000.0 1.884985006925 3 16 18 12500000.0 1.90771500 12600000.0 1.93044500 12700000.0 1.953175006925 3 16 19 12800000.0 1.97590500 12900000.0 1.99863500 13000000.0 2.009486676925 3 16 20 13100000.0 2.00846000 13200000.0 2.00743333 13300000.0 2.006406676925 3 16 21 13400000.0 2.00538000 13500000.0 2.00435333 13600000.0 2.003326676925 3 16 22 13700000.0 2.00230000 13800000.0 2.00127333 13900000.0 2.000246676925 3 16 23 14000000.0 1.99922000 14100000.0 1.99819333 14200000.0 1.997166676925 3 16 24 14300000.0 1.99614000 14400000.0 1.99511333 14500000.0 1.994686676925 3 16 25 14600000.0 1.99486000 14700000.0 1.99503333 14800000.0 1.995206676925 3 16 26 14900000.0 1.99538000 15000000.0 1.99555333 15100000.0 1.995726676925 3 16 27 15200000.0 1.99590000 15300000.0 1.99607333 15400000.0 1.996246676925 3 16 28 15500000.0 1.99642000 15600000.0 1.99659333 15700000.0 1.996766676925 3 16 29 15800000.0 1.99694000 15900000.0 1.99711333 16000000.0 1.994380006925 3 16 30 16100000.0 1.98874000 16200000.0 1.98310000 16300000.0 1.977460006925 3 16 31 16400000.0 1.97182000 16500000.0 1.96618000 16600000.0 1.960540006925 3 16 32 16700000.0 1.95490000 16800000.0 1.94926000 16900000.0 1.943620006925 3 16 33 17000000.0 1.91730500 17100000.0 1.87031500 17200000.0 1.823325006925 3 16 34 17300000.0 1.77633500 17400000.0 1.72934500 17500000.0 1.682355006925 3 16 35 17600000.0 1.63536500 17700000.0 1.58837500 17800000.0 1.541385006925 3 16 36 17900000.0 1.49439500 18000000.0 1.45198500 18100000.0 1.414155006925 3 16 37 18200000.0 1.37632500 18300000.0 1.33849500 18400000.0 1.300665006925 3 16 38 18500000.0 1.26283500 18600000.0 1.22500500 18700000.0 1.187175006925 3 16 39 18800000.0 1.14934500 18900000.0 1.11151500 19000000.0 1.081435006925 3 16 40 19100000.0 1.05910500 19200000.0 1.03677500 19300000.0 1.014445006925 3 16 41 19400000.0 .992115000 19500000.0 .969785000 19600000.0 .9474550006925 3 16 42 19700000.0 .925125000 19800000.0 .902795000 19900000.0 .8804650006925 3 16 43 20000000.0 0.0 6925 3 16 44 6925 3 099999 6925 0 0 0 6.91690E+4 1.67483E+2 0 0 0 1692533 16 1 0.000000+0 0.000000+0 0 16 0 1692533 16 2 0.000000+0 0.000000+0 1 5 120 15692533 16 3 1.000000-5 8.000000+6 8.240180+6 8.700000+6 9.500000+6 1.050000+7692533 16 4 1.150000+7 1.250000+7 1.375000+7 1.525000+7 1.650000+7 1.750000+7692533 16 5 1.850000+7 1.950000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0692533 16 6 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0692533 16 7 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 1.544490-3692533 16 8 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0692533 16 9 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0692533 16 10 1.544490-3 2.161260-4 8.021960-5 8.018380-5 8.965270-5 1.308120-4692533 16 11 8.611730-5 7.978640-5 1.636990-4 1.667380-4 9.654600-5 2.752760-4692533 16 12 1.762320-3 1.773580-4 1.789170-4 1.261310-4 1.126210-4 7.632130-5692533 16 13 6.414960-5 1.179690-4 1.202010-4 7.090170-5 1.789260-4 2.253400-3692533 16 14 3.400620-4 1.941300-4 1.014070-4 7.258540-5 4.974090-5 6.811760-5692533 16 15 7.143290-5 4.130170-5 6.744160-5 2.055720-3 1.967720-4 1.028640-4692533 16 16 7.314980-5 5.047840-5 6.867560-5 7.201810-5 4.176900-5 6.741150-5692533 16 17 7.739520-4 1.059610-4 7.502940-5 7.712840-5 9.702920-5 1.017520-4692533 16 18 5.552930-5 7.629130-5 2.742330-4 8.094930-5 8.206160-5 1.132040-4692533 16 19 1.180210-4 6.526070-5 1.110700-4 1.449620-4 5.756050-5 7.774620-5692533 16 20 8.052350-5 5.484240-5 7.438890-5 4.765490-4 1.022420-4 1.031120-4692533 16 21 1.117260-4 7.429010-5 3.972050-4 1.424560-4 1.295420-4 1.435500-4692533 16 22 4.368100-4 1.091050-4 1.436310-4 2.619390-3 1.132960-4 1.091640-3692533 16 23 692533 099999 6925 0 0 0 0 0 0 0 7.31810E+4 1.79393E+2 0 0 34 107328 1451 1 0.0 0.0 0 0 0 67328 1451 2 1.00000E+0 2.00000E+7 0 0 10 20027328 1451 3 3.00000E+2 0.0 1 0 117 47328 1451 4 73-Ta-181 NAIG EVAL-MAR87 N.YAMAMURO 7328 1451 5 DIST-Feb2004 7328 1451 6 ----IRDF-2002 MATERIAL 7328 REVISION 2 7328 1451 7 -----INCIDENT NEUTRON DATA 7328 1451 8 ------ENDF-6 FORMAT 7328 1451 9 **************************************************************** 7328 1451 10 73-TA-181 NAIG EVAL-MAR87 N.YAMAMURO 7328 1451 11 DIST-SEP89 REV2-FEB94 7328 1451 12 ----JENDL-3.2 MATERIAL 7328 REVISION 2 7328 1451 13 7328 1451 14 TA-181 CAPTURE TA-181(HALF-LIFE = 114.43D) 7328 1451 15 **************************************************************** 7328 1451 16 HISTORY 7328 1451 17 76-03 THE EVALUATION FOR JENDL-1 /1/ WAS MADE BY H.YAMAKOSHI 7328 1451 18 (SHIP RESEARCH INSTITUTE) AND JENDL-1 COMPILATION GROUP. 7328 1451 19 83-03 JENDL-1 DATA WERE ADOPTED FOR JENDL-2 AND EXTENDED TO 20 7328 1451 20 MEV. MF=5 WAS REVISED, AND UNRESOLVED RESONANCE PARAMETERS 7328 1451 21 WERE ADDED BY Y.KIKUCHI (JAERI) /2/. 7328 1451 22 83-11 COMMENT DATA WERE ADDED. 7328 1451 23 87-03 THE EVALUATION FOR JENDL-3 WAS MADE BY N.YAMAMURO (NAIG). 7328 1451 24 RESONANCE PARAMETERS WERE ADDED BY NEW EXPERIMENTAL DATA. 7328 1451 25 NEUTRON CROSS SECTIONS, EXCEPT TOTAL AND ELASTIC SCATTERING 7328 1451 26 CROSS SECTIONS, AND ENERGY DISTRIBUTIONS OF SECONDARY 7328 1451 27 NEUTRONS AND PHOTONS WERE CALCULATED WITH GNASH /3/ AND 7328 1451 28 CASTHY /4/ CODES. 7328 1451 29 94-02 JENDL-3.2. 7328 1451 30 COMPILED BY T.NAKAGAWA (NDC/JAERI) 7328 1451 31 7328 1451 32 ***** MODIFIED PARTS FOR JENDL-3.2 ******************** 7328 1451 33 (12,102) 7328 1451 34 DATA WERE DETERMINED FROM ENERGY BALANCE. 7328 1451 35 *********************************************************** 7328 1451 36 7328 1451 37 7328 1451 38 MF=1 GENERAL INFORMATION 7328 1451 39 MT=451 DESCRIPTIVE DATA AND DICTIONARY 7328 1451 40 7328 1451 41 MF=2 RESONANCE PARAMETERS 7328 1451 42 MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETERS 7328 1451 43 RESOLVED PARAMETERS FOR MLBW FORMULA 7328 1451 44 7328 1451 45 THE ENERGY REGION FROM 1.0E-5EV TO 1.0 KEV. PARAMETERS WERE 7328 1451 46 TAKEN FROM REFS./5,6,7/ FOR POSITIVE RESONANCES, AND FROM 7328 1451 47 ENDF/B-IV FOR A NEGATIVE RESONANCE. THE RADIATIVE WIDTH OF 7328 1451 48 0.059EV WAS ASSUMED FOR THE RESONANCE WHOSE RADIATIVE WIDTH 7328 1451 49 WAS UNKNOWN. 7328 1451 50 7328 1451 51 UNRESOLVED PARAMETERS 7328 1451 52 7328 1451 53 IN THE ENERGY RANGE FROM 1 TO 100KEV, PARAMETERS WERE 7328 1451 54 DETERMINED TO REPRODUCE THE MEASURED CAPTURE CROSS SECTIONS 7328 1451 55 /6,8/. THE PARAMETERS ARE AS FOLLOWS, 7328 1451 56 7328 1451 57 R= 7.8 FM , DOBS= 4.2 EV , RADIATIVE WIDTH= 0.065 EV, 7328 1451 58 SO= 1.7E-04 S1= 2.0E-05 S2= 2.3E-04 NL= 3 7328 1451 59 7328 1451 60 CALCULATED 2200-M/SEC CROSS SECTIONS AND RESONANCE INTEGRALS7328 1451 61 2200-M/SEC RES. INTEG 7328 1451 62 ELASTIC 5.65 B - 7328 1451 63 CAPTURE 20.67 B 660.43 B 7328 1451 64 TOTAL 26.32 B - 7328 1451 65 7328 1451 66 MF=3 NEUTRON CROSS SECTIONS 7328 1451 67 MT=1 TOTAL 7328 1451 68 EVALUATED FROM EXPERIMENTAL DATA. 7328 1451 69 MT=2 ELASTIC SCATTERING 7328 1451 70 (TOTAL CROSS SECTION) - (REACTION CROSS SECTION) 7328 1451 71 MT=102 RADIATIVE CAPTURE CROSS SECTION 7328 1451 72 CALCULATED WITH CASTHY/4/. 7328 1451 73 7328 1451 74 REFERENCES 7328 1451 75 1) IGARASI S. ET AL.: JAERI 1261 (1979) 7328 1451 76 2) (ED.) NAKAGAWA T.: JAERI-M 84-103 (1984) 7328 1451 77 3) YOUNG,P.G. AND ARTHUR,E.D.: "GNASH, A PREEQUILIBRIUM 7328 1451 78 STATISTICAL NUCLEAR-MODEL CODE FOR CALCULATION OF CROSS 7328 1451 79 SECTIONS AND EMISSION SPECTRA", LA-6974 (1977). 7328 1451 80 4) IGARASHI,S. AND FUKAHORI,T.: JAERI 1321 (1991). 7328 1451 81 5) MUGHABGHAB,S.F, AND GARDER,D.I.: BNL325, 3RD ED. (1973). 7328 1451 82 6) MACKLIN,R,L,: NUCL,SCI,ENG., 86,362 (1984). 7328 1451 83 7) TSUBONE,I., NAKAJIMA,Y. AND KANDA,Y.: PRIVATE COMMUNICATION 7328 1451 84 8) YAMAMURO,N., SAITO,K., EMOTO,T., WADA,T., FUJITA,Y. AND 7328 1451 85 KOBAYASHI,K.: J.NUCL.SCI.TECHNOL., 17, 582 (1980). 7328 1451 86 9) WILMORE,D. AND HODGSON,P.E.: NUCL, PHYS., 55, 673 (1964). 7328 1451 87 10) FIRESTONE,R.B.* NUCL, DATA SHEETS 43, 289 (1984). 7328 1451 88 *************************************************************** 7328 1451 89 The Q values and threshold energies were updated prior to pro- 7328 1451 90 cessing through the codes to comply with the values obtained 7328 1451 91 using the NNDC calculation program which is based on the 1995 7328 1451 92 Update to the Atomic mass Evaluation. 7328 1451 93 ***************** Program LINEAR (VERSION 2002-1) ***************7328 1451 94 For All Data Greater than 1.0000E-10 barns in Absolute Value 7328 1451 95 Data Linearized to Within an Accuracy of .100000000 per-cent 7328 1451 96 ***************** Program RECENT (VERSION 2002-1) ***************7328 1451 97 for All Data Greater than 1.0000E-10 barns in Absolute Value 7328 1451 98 Data Linearized to within an Accuracy of .100000000 per-cent 7328 1451 99 ***************** Program SIGMA1 (VERSION 2002-1) ***************7328 1451 100 Data Doppler Broadened to 300.000000 Kelvin 7328 1451 101 for All Data Greater than 1.0000E-10 barns in Absolute Value 7328 1451 102 Data Linearized to Within an Accuracy pf .100000000 per-cent 7328 1451 103 ***************** Program FIXUP (Version 2002-1) ****************7328 1451 104 Corrected ZA/AWR in All Sections-----------------------------Yes 7328 1451 105 Corrected Thresholds-----------------------------------------Yes 7328 1451 106 Extended Cross Sections to 20 MeV----------------------------No 7328 1451 107 Allow Cross Section Deletion---------------------------------No 7328 1451 108 Allow Cross Section Reconstruction---------------------------No 7328 1451 109 Make All Cross Sections Non-Negative-------------------------Yes 7328 1451 110 Delete Energies Not in Ascending Order-----------------------Yes 7328 1451 111 Deleted Duplicate Points-------------------------------------Yes 7328 1451 112 Check for Ascending MAT/MF/MT Order--------------------------Yes 7328 1451 113 Check for Legal MF/MT Numbers--------------------------------Yes 7328 1451 114 Allow Creation of Missing Sections---------------------------No 7328 1451 115 Allow Insertion of Energy Points-----------------------------No 7328 1451 116 Create Uniform Energy Grid-----------------------------------No 7328 1451 117 Delete Section if Cross Section =0 at All Energies-----------Yes 7328 1451 118 ***************** Program GROUPIE (VERSION 2002-1) **************7328 1451 119 Unshielded Group Averages Using 640 Groups 7328 1451 120 Weighting Spectrum: Flat (Constant) Spectrum 7328 1451 121 1 451 125 17328 1451 122 2 151 4 17328 1451 123 3 102 217 17328 1451 124 33 102 16 17328 1451 125 7328 1 099999 7328 0 0 0 7.31810E+4 1.79394E+2 0 0 1 07328 2151 1 7.31810E+4 1.00000E+0 0 0 1 07328 2151 2 1.00000E-5 2.40000E+3 0 0 0 07328 2151 3 3.50000E+0 7.80000E-1 0 0 0 07328 2151 4 7328 2 099999 7328 0 0 0 7.31810E+4 1.79393E+2 0 0 0 07328 3102 1 6.06296E+6 6.06296E+6 0 0 1 6417328 3102 2 641 1 7328 3102 3 .000100000 323.217903 .000105000 315.681488 .000110000 308.5218207328 3102 4 .000115000 301.924447 .000120000 294.211880 .000127500 285.6549077328 3102 5 .000135000 277.837071 .000142500 270.652682 .000150000 262.8772027328 3102 6 .000160000 254.777193 .000170000 247.407203 .000180000 240.6421387328 3102 7 .000190000 234.347828 .000200000 228.601006 .000210000 223.1942217328 3102 8 .000220000 218.172206 .000230000 213.506120 .000240000 208.0290837328 3102 9 .000255000 202.010625 .000270000 197.366526 .000280000 192.1967877328 3102 10 .000300000 185.892130 .000320000 180.183474 .000340000 174.9552667328 3102 11 .000360000 170.139778 .000380000 165.743885 .000400000 161.1609207328 3102 12 .000425000 156.471122 .000450000 152.185981 .000475000 148.2575327328 3102 13 .000500000 144.563214 .000525000 141.186770 .000550000 137.9971507328 3102 14 .000575000 135.030648 .000600000 131.990183 .000630000 128.8623417328 3102 15 .000660000 126.000351 .000690000 123.254902 .000720000 120.3335897328 3102 16 .000760000 117.201272 .000800000 114.293003 .000840000 111.6277757328 3102 17 .000880000 109.094632 .000920000 106.772417 .000960000 104.5484537328 3102 18 .001000000 102.243220 .001050000 99.8368250 .001100000 97.58099947328 3102 19 .001150000 95.5024452 .001200000 93.0505503 .001275000 90.35585597328 3102 20 .001350000 87.8922920 .001425000 85.5932451 .001500000 83.15089677328 3102 21 .001600000 80.5973361 .001700000 78.2663165 .001800000 76.11048847328 3102 22 .001900000 74.1411257 .002000000 72.3150953 .002100000 70.60153967328 3102 23 .002200000 69.0322893 .002300000 67.5316982 .002400000 65.81944087328 3102 24 .002550000 63.9166791 .002700000 62.4282767 .002800000 60.81092297328 3102 25 .003000000 58.8199465 .003200000 57.0083580 .003400000 55.35215557328 3102 26 .003600000 53.8422601 .003800000 52.4486161 .004000000 50.99654757328 3102 27 .004250000 49.5228933 .004500000 48.1779215 .004750000 46.92074987328 3102 28 .005000000 45.7666443 .005250000 44.6962028 .005500000 43.68386187328 3102 29 .005750000 42.7554179 .006000000 41.7800919 .006300000 40.80494017328 3102 30 .006600000 39.8823962 .006900000 39.0272497 .007200000 38.09835167328 3102 31 .007600000 37.1060107 .008000000 36.1997943 .008400000 35.35036067328 3102 32 .008800000 34.5611280 .009200000 33.8269194 .009600000 33.12773187328 3102 33 .010000000 32.4026932 .010500000 31.6356982 .011000000 30.93437197328 3102 34 .011500000 30.2654520 .012000000 29.4991382 .012750000 28.65129477328 3102 35 .013500000 27.8663681 .014250000 27.1466861 .015000000 26.37937077328 3102 36 .016000001 25.5739361 .017000001 24.8352752 .017999999 24.16073867328 3102 37 .018999999 23.5405789 .020000000 22.9587048 .021000000 22.42625467328 3102 38 .022000000 21.9247118 .023000000 21.4575784 .024000000 20.90912817328 3102 39 .025500000 20.3185917 .027000001 19.8510914 .028000001 19.34385857328 3102 40 .029999999 18.7191818 .032000002 18.1495657 .034000002 17.62914457328 3102 41 .035999998 17.1541449 .037999999 16.7162398 .039999999 16.25956907328 3102 42 .042500000 15.7963649 .045000002 15.3734911 .047499999 14.98001797328 3102 43 .050000001 14.6188762 .052499998 14.2855256 .055000000 13.97040047328 3102 44 .057500001 13.6816926 .059999999 13.3779349 .063000001 13.07361657328 3102 45 .066000000 12.7868579 .068999998 12.5198571 .071999997 12.23061137328 3102 46 .075999998 11.9210685 .079999998 11.6386548 .083999999 11.37346237328 3102 47 .088000000 11.1271550 .092000000 10.8993591 .096000001 10.68200867328 3102 48 .100000001 10.4585554 .104999997 10.2230982 .109999999 10.01148777328 3102 49 .115000002 9.80350319 .119999997 9.56591573 .127499998 9.302968457328 3102 50 .135000005 9.06519879 .142499998 8.85047362 .150000006 8.611181927328 3102 51 .159999996 8.36500915 .170000002 8.14190890 .180000007 7.936169207328 3102 52 .189999998 7.74863484 .200000003 7.57624267 .209999993 7.414342437328 3102 53 .219999999 7.26714837 .230000004 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2.000000+7 0.000000+0 732833102 16 732833 099999 7328 0 0 0 0 0 0 0 7.41860E+4 1.84357E+2 0 0 34 107443 1451 1 0.0 0.0 0 0 0 67443 1451 2 1.00000E+0 2.00000E+7 0 0 10 20027443 1451 3 3.00000E+2 0.0 1 0 199 47443 1451 4 74-W -186 FEI,LANL+ EVAL-SEP01 K.I.Zolotarev,M.B.CHADWICK etal. 7443 1451 5 DIST-Feb2004 7443 1451 6 ----IRDF-2002 MATERIAL 7443 7443 1451 7 -----INCIDENT NEUTRON DATA 7443 1451 8 ------ENDF-6 FORMAT 7443 1451 9 ***************************************************************** 7443 1451 10 74-W -186 FEI EVAL-Sep01 K.I.Zolotarev 7443 1451 11 DIST-Jan04 7443 1451 12 ----BROND-3 MATERIAL 7443 Revision 1, Jan. 2004 7443 1451 13 -----INCIDENT NEUTRON DATA 7443 1451 14 ------ENDF-6 FORMAT 7443 1451 15 ------Russian Reactor Dosimetry File RRDF-2002 7443 1451 16 ***************************************************************** 7443 1451 17 ********** Start of (N,G) bibliographical component ********** 7443 1451 18 ***************************************************************** 7443 1451 19 ***************************************************************** 7443 1451 20 Authors of evaluation: K.I.Zolotarev and V.G.Pronyaev 7443 1451 21 ***************************************************************** 7443 1451 22 7443 1451 23 ----- MF=2 MT=151 ----- 7443 1451 24 7443 1451 25 RESOLVED RESONANCE PARAMETERS 7443 1451 26 Resolved MLBW resonance parameters up to 8.5 keV were based 7443 1451 27 on data from ref.[1-2]. Spins of resonances, which were not iden- 7443 1451 28 tified, have been assigned by taking into account 2*J+1 law and 7443 1451 29 approximate linear growth of the number of levels with given spin 7443 1451 30 and parity from energy. Average gamma-widths 0.05 eV (L=0 and L=1)7443 1451 31 were assigned to those resonances, widths for which were absent. 7443 1451 32 Fictitious resonances were added in the energy region 4 - 8.5 keV 7443 1451 33 to obtain the average capture cross section values observed in 7443 1451 34 the experiments [3],[4],[5],[6]. 7443 1451 35 7443 1451 36 UNRESOLVED RESONANCE PARAMETERS 7443 1451 37 Unresolved resonance parameters in the region 8.5 - 100 keV 7443 1451 38 were evaluated with using 'EVPAR' code (Hauser-Feshbach-Moldauer 7443 1451 39 statistical model)[7]. They were calculated from optimized fit to 7443 1451 40 the measured capture [4,5,6,8,9] and total [10,11] cross sections 7443 1451 41 data. Average resolved resonance parameters have been used as 7443 1451 42 zero approximation. Contribution of F-wave in the energy region 7443 1451 43 30 - 100 keV given as smoothed background cross section in File 3.7443 1451 44 7443 1451 45 ----- MF=3 MT=102 ----- 7443 1451 46 Capture cross sections from 1.000E-5 eV to 100 keV are recon- 7443 1451 47 structed from evaluated MLBW resolved and unresolved resonance 7443 1451 48 parameters. Contribution of F-wave in the neutron energy interval 7443 1451 49 30 - 100 keV is added as described above. 7443 1451 50 Data base for the evaluation W-186(n,g)W-187 excitation func- 7443 1451 51 tion in the neutron energy range 100 keV - 20 MeV was formed from 7443 1451 52 microscopic experimental data [4-6],[8-9],[12-25] and data recei- 7443 1451 53 ved from theoretical model calculations. Experimental data [8-9], 7443 1451 54 [16-20], [22], [24] were corrected to the new standards. Uncer- 7443 1451 55 tainty for cross section values measured by Bartolome et al. [4] 7443 1451 56 was evaluated as +-16%. Theoretical model calculations was done 7443 1451 57 by means of GNASH code [26]. New modified version of GNASH was 7443 1451 58 utilized for calculations [27]. Data from GNASH calculation were 7443 1451 59 dominant in the neutron energy range 4 - 20 MeV. 7443 1451 60 The evaluation W-186(n,g)W-187 excitation function in the 7443 1451 61 energy range above 100 keV has been carried out within the frame- 7443 1451 62 work of generalized least squares method , rational function was 7443 1451 63 used as model function [28]. Procedure of calculation recommended 7443 1451 64 cross section data was performed by means of PADE-2 code [29]. 7443 1451 65 Capture cross section at En=0.0253 eV and Resonance Integral 7443 1451 66 (0.5 to 2.0E+7 eV) from present evaluation are given below in the 7443 1451 67 Table 1. in a comparison with data from compilations [1] and [30].7443 1451 68 Table 1. 7443 1451 69 -------------------------------------------------- 7443 1451 70 Parameters Values, barn References 7443 1451 71 -------------------------------------------------- 7443 1451 72 Capture cross section 38.47 +- 0.85 This eval. 7443 1451 73 at En=0.0253 eV 37.9 +- 0.6 [ 1] 7443 1451 74 37 +- 2 [30] 7443 1451 75 38.5 +- 0.5 [31] 7443 1451 76 -------------------------------------------------- 7443 1451 77 Resonance Integral 482.2 This eval. 7443 1451 78 485 +- 15 [ 1] 7443 1451 79 485 +- 15 [30] 7443 1451 80 485 +- 15 [31] 7443 1451 81 -------------------------------------------------- 7443 1451 82 7443 1451 83 ----- MF=33 MT=102----- 7443 1451 84 Uncertainties in the evaluated excitation function for the 7443 1451 85 reaction W-186(n,g)W-187 are given in the three independent 7443 1451 86 matrixes. 7443 1451 87 In the energy range 1.000E-05 - 100 eV uncertainties are pre- 7443 1451 88 sented in the form of relative covariance matrix for the 14-neut- 7443 1451 89 ron energy groups (LB=5). Eigenvalues of this 6-th digits rela- 7443 1451 90 tive covariance matrix are the following: 7443 1451 91 7443 1451 92 1.14625E-10 1.09510E-07 7.28111E-07 8.61003E-06 7443 1451 93 4.80743E-05 5.34800E-05 5.91642E-05 6.60535E-05 7443 1451 94 8.44288E-05 1.48138E-04 1.95936E-04 2.39607E-04 7443 1451 95 8.15874E-04 8.38515E-03 7443 1451 96 7443 1451 97 In the energy range 0.1 - 8.5 keV uncertainties are given in 7443 1451 98 the form of diagonal matrix of uncertainties for 19-th neutron 7443 1451 99 energy intervals (LB=1) 7443 1451 100 In the energy range 8.5 keV - 20 MeV uncertainties are pre- 7443 1451 101 sented in the form of relative covariance matrix for the 24-neut- 7443 1451 102 ron energy groups (LB=5). Eigenvalues of this 6-th digits rela- 7443 1451 103 tive covariance matrix are the following: 7443 1451 104 7443 1451 105 6.81049E-10 1.45633E-05 5.83401E-05 1.88615E-04 7443 1451 106 6.00477E-04 7.08017E-04 9.54276E-04 1.08032E-03 7443 1451 107 1.26144E-03 1.35976E-03 1.50022E-03 1.66258E-03 7443 1451 108 1.66943E-03 2.15756E-03 2.87259E-03 3.83276E-03 7443 1451 109 6.81420E-03 9.70769E-03 9.99331E-03 1.17664E-02 7443 1451 110 1.48009E-02 1.93593E-02 2.04003E-02 2.26206E-01 7443 1451 111 7443 1451 112 References : 7443 1451 113 1. S.F.Mughabghab et al. Neutron Cross Sections, vol.1, part B, 7443 1451 114 New York, Academic Press, 1984 7443 1451 115 2. S.I.Sukhoruchkin et al. Landolt Bornstein New Series, v.I/16B,7443 1451 116 ed. H.Schopper, Springer, 1998, pp.w19-w23 7443 1451 117 3. S.V.Kapchigashev, Ju.P.Popov, F.L.Shapiro EXFOR 40755.005 7443 1451 118 4. Z.M.Bartolome et al. Nucl. Sci. Eng., v.37, pp.137-156, 1969 7443 1451 119 5. R.L.Macklin et al. Nucl. Sci. Eng., v.84, p.98, June 1983 7443 1451 120 6. M.V.Bokhovko et al. Voprosy Atomnoj Nauki i Tekhniki, Ser.: 7443 1451 121 Yadernye Konstanty, v.1, p.39, February 1986 7443 1451 122 7. G.N.Manturov et al. Voprosy Atomnoy Nauki i Tekhniki, Ser.: 7443 1451 123 Jadernye Konstanty, v.1, p.50, 1983 7443 1451 124 8. Yu.Ya.Stavisskii, V.A.Tolstikov Atomnaja Energija, v.9, 7443 1451 125 p.401, 1960 7443 1451 126 9. V.N.Kononov et al. Yadernaja Fizika, v.4, no.2, p.282, 1966 7443 1451 127 10. P.T.Guenther. et al. Phys. Rev. C., v.26, p.2433, 1982 7443 1451 128 11. V.N.Kononov et al. Sov J. of Nuclear Physics, v.46, part 1, 7443 1451 129 p.51, July 1987 7443 1451 130 12. R.Allen et al. Nature, v.161, p.727, May 1948 7443 1451 131 13. L.E.Beghian, H.H.Halban Nature, v.163, p.366, March 1949 7443 1451 132 14. J.L.Perkin et al. Proc. Phys. Soc., v.72, p.505, 1958 7443 1451 133 15. A.I.Leipunskij et al. Second UN Conf. on the Peaceful Uses of 7443 1451 134 Atomic Energy, Geneva, 1-13 September 1958, v.15, p.50(2219) 7443 1451 135 16. M.V.Pasechnik et al. Second UN Conf. on the Peaceful Uses of 7443 1451 136 Atomic Energy, Geneva, 1-13 September 1958, v.15, p.18(2030) 7443 1451 137 17. A.E.Johnsrud et al. Phys. Rev., v.116, p.927, 1959 7443 1451 138 18. G.G.Zaikin et al. Jaderno-Fizicheskie Issledovanija (USSR 7443 1451 139 progress report), v.6, p.103, 1968 7443 1451 140 19. M.Lindner et al. Nucl. Sci. Eng., v.59, p.381, April 1976 7443 1451 141 20. O.Schwerer et al. Nucl. Phys., v.A264, p.105, June 1976 7443 1451 142 21. M.Valkonen Report JU-RR-1/1976, Jyvaeskylae University, 1976 7443 1451 143 M.Valkonen et al. J. Inorg. Nucl. Chem., v.36, p.715, 1974 7443 1451 144 22. G.Magnusson et al. Physica Scripta, v.21, p.21, January 1980 7443 1451 145 23. J.Voignier et al. Report CEA-R-5089, August 1981 ; 7443 1451 146 J.Voignier et al. Nucl. Sci. Eng., v.93, p.43, 1986 7443 1451 147 24. Yu.N.Trofimov 1st International Conf. on Neutron Physics, 7443 1451 148 Kiev, 14-16 September 1987, v.3, p.331 7443 1451 149 25. Zhang Guohui, Lu Hanlin et al. Communication of Nuclear Data 7443 1451 150 Progress, No. 23, Atomic Energy Press, Beijing, June, 2000 7443 1451 151 26. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 7443 1451 152 Model Code for Calculation of Cross Section and Emission 7443 1451 153 Spectra. Report LA-6947, Los Alamos, 1977 7443 1451 154 27. E.L.Trykov, G.Ya.Tertychnyi Private communication, IPPE, 7443 1451 155 Obninsk, May 1999 7443 1451 156 28. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 7443 1451 157 st's Meeting on Evaluation and Processing of Covariance Data, 7443 1451 158 Oak Ridge, USA, 7-9 September 1992, OECD, Paris, 1993, p.105 7443 1451 159 29. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 7443 1451 160 30. A.V.Ignatyuk et al. Landolt Bornstein New Series, v.I/16A, 7443 1451 161 Part 1, ed. H.Schopper, Springer, 1998, p.8-24 7443 1451 162 31. S.F.Mughabghab et al. Termal Neutron Capture Cross Sections 7443 1451 163 Resonance Integrals and G-factors, Report INDC(NDS)-440, 7443 1451 164 IAEA, Vienna, February 2003 7443 1451 165 ***************************************************************** 7443 1451 166 ********** End of (N,G) bibliographical component ********** 7443 1451 167 ***************************************************************** 7443 1451 168 74-W -186 LANL,ANL EVAL-OCT96 M.B.CHADWICK,P.G.YOUNG,ARTHUR 7443 1451 169 Ch96a,Ch96b,Ar80,Ch99DIST-SEP 1 REV1- 20010926 7443 1451 170 ----ENDF/B-VI MATERIAL 7443 7443 1451 171 MF=3 MT= 1 Total Cross Section 7443 1451 172 MT= 2 Elastic Scattering Cross Section 7443 1451 173 ***************************************************************** 7443 1451 174 7443 1451 175 ***************** Program LINEAR (VERSION 2002-1) ***************7443 1451 176 For All Data Greater than 1.0000E-10 barns in Absolute Value 7443 1451 177 Data Linearized to Within an Accuracy of .100000000 per-cent 7443 1451 178 ***************** Program RECENT (VERSION 2002-1) ***************7443 1451 179 for All Data Greater than 1.0000E-10 barns in Absolute Value 7443 1451 180 Data Linearized to within an Accuracy of .100000000 per-cent 7443 1451 181 ***************** Program SIGMA1 (VERSION 2002-1) ***************7443 1451 182 Data Doppler Broadened to 300.000000 Kelvin 7443 1451 183 for All Data Greater than 1.0000E-10 barns in Absolute Value 7443 1451 184 Data Linearized to Within an Accuracy pf .100000000 per-cent 7443 1451 185 ***************** Program FIXUP (Version 2002-1) ****************7443 1451 186 Corrected ZA/AWR in All Sections-----------------------------Yes 7443 1451 187 Corrected Thresholds-----------------------------------------Yes 7443 1451 188 Extended Cross Sections to 20 MeV----------------------------No 7443 1451 189 Allow Cross Section Deletion---------------------------------No 7443 1451 190 Allow Cross Section Reconstruction---------------------------No 7443 1451 191 Make All Cross Sections Non-Negative-------------------------Yes 7443 1451 192 Delete Energies Not in Ascending Order-----------------------Yes 7443 1451 193 Deleted Duplicate Points-------------------------------------Yes 7443 1451 194 Check for Ascending MAT/MF/MT Order--------------------------Yes 7443 1451 195 Check for Legal MF/MT Numbers--------------------------------Yes 7443 1451 196 Allow Creation of Missing Sections---------------------------No 7443 1451 197 Allow Insertion of Energy Points-----------------------------No 7443 1451 198 Create Uniform Energy Grid-----------------------------------No 7443 1451 199 Delete Section if Cross Section =0 at All Energies-----------Yes 7443 1451 200 ***************** Program GROUPIE (VERSION 2002-1) **************7443 1451 201 Unshielded Group Averages Using 640 Groups 7443 1451 202 Weighting Spectrum: Flat (Constant) Spectrum 7443 1451 203 1 451 207 17443 1451 204 2 151 4 17443 1451 205 3 102 217 17443 1451 206 33 102 94 17443 1451 207 7443 1 099999 7443 0 0 0 7.41860E+4 1.84357E+2 0 0 1 07443 2151 1 7.41860E+4 1.00000E+0 0 0 1 07443 2151 2 1.00000E-5 8.50000E+3 0 0 0 07443 2151 3 0.0 7.64000E-1 0 0 0 07443 2151 4 7443 2 099999 7443 0 0 0 7.41860E+4 1.84357E+2 0 0 0 07443 3102 1 5.46700E+6 5.46700E+6 0 0 1 6417443 3102 2 641 1 7443 3102 3 .000100000 603.335846 .000105000 589.267631 .000110000 575.9026837443 3102 4 .000115000 563.587364 .000120000 549.190298 .000127500 533.2169627443 3102 5 .000135000 518.623372 .000142500 505.212229 .000150000 490.6976817443 3102 6 .000160000 475.577316 .000170000 461.819668 .000180000 449.1912287443 3102 7 .000190000 437.441542 .000200000 426.713841 .000210000 416.6208847443 3102 8 .000220000 407.246173 .000230000 398.535864 .000240000 388.3117177443 3102 9 .000255000 377.076879 .000270000 368.407585 .000280000 358.7570467443 3102 10 .000300000 346.987888 .000320000 336.331282 .000340000 326.5715237443 3102 11 .000360000 317.582189 .000380000 309.376115 .000400000 300.8208007443 3102 12 .000425000 292.066023 .000450000 284.066608 .000475000 276.7330497443 3102 13 .000500000 269.836551 .000525000 263.533426 .000550000 257.5790457443 3102 14 .000575000 252.041159 .000600000 246.365181 .000630000 240.5260707443 3102 15 .000660000 235.183230 .000690000 230.057942 .000720000 224.6043127443 3102 16 .000760000 218.756748 .000800000 213.327421 .000840000 208.3517907443 3102 17 .000880000 203.622726 .000920000 199.287405 .000960000 195.1354907443 3102 18 .001000000 190.831817 .001050000 186.339264 .001100000 182.1277957443 3102 19 .001150000 178.247247 .001200000 173.669665 .001275000 168.6387387443 3102 20 .001350000 164.039287 .001425000 159.746960 .001500000 155.1870537443 3102 21 .001600000 150.419482 .001700000 146.067368 .001800000 142.0423177443 3102 22 .001900000 138.365382 .002000000 134.956017 .002100000 131.7566167443 3102 23 .002200000 128.826606 .002300000 126.024764 .002400000 122.8276247443 3102 24 .002550000 119.274676 .002700000 116.495377 .002800000 113.4751727443 3102 25 .003000000 109.757167 .003200000 106.374065 .003400000 103.2810967443 3102 26 .003600000 100.461353 .003800000 97.8587164 .004000000 95.14697857443 3102 27 .004250000 92.3949345 .004500000 89.8831489 .004750000 87.53526427443 3102 28 .005000000 85.3797497 .005250000 83.3803562 .005500000 81.48937897443 3102 29 .005750000 79.7549491 .006000000 77.9328522 .006300000 76.11093107443 3102 30 .006600000 74.3872531 .006900000 72.7894516 .007200000 71.05385907443 3102 31 .007600000 69.1997423 .008000000 67.5065760 .008400000 65.92103847443 3102 32 .008800000 64.4499949 .009200000 63.0772891 .009600000 61.76737067443 3102 33 .010000000 60.4112555 .010500000 58.9776762 .011000000 57.66673047443 3102 34 .011500000 56.4163054 .012000000 54.9836997 .012750000 53.39855427443 3102 35 .013500000 51.9309308 .014250000 50.5851860 .015000000 49.15023247443 3102 36 .016000001 47.6438249 .017000001 46.2621264 .017999999 45.00019547443 3102 37 .018999999 43.8399023 .020000000 42.7512128 .021000000 41.75505657443 3102 38 .022000000 40.8167414 .023000000 39.9428153 .024000000 38.91634697443 3102 39 .025500000 37.8095093 .027000001 36.9325200 .028000001 35.97969527443 3102 40 .029999999 34.8066273 .032000002 33.7378775 .034000002 32.76244257443 3102 41 .035999998 31.8726360 .037999999 31.0529462 .039999999 30.19880417443 3102 42 .042500000 29.3331880 .045000002 28.5416236 .047499999 27.80334387443 3102 43 .050000001 27.1230983 .052499998 26.4935146 .055000000 25.89745037443 3102 44 .057500001 25.3521109 .059999999 24.7788712 .063000001 24.20585197443 3102 45 .066000000 23.6660150 .068999998 23.1633318 .071999997 22.61846327443 3102 46 .075999998 22.0351607 .079999998 21.5031559 .083999999 21.00417257443 3102 47 .088000000 20.5414942 .092000000 20.1120657 .096000001 19.70136977443 3102 48 .100000001 19.2765995 .104999997 18.8267371 .109999999 18.41600767443 3102 49 .115000002 18.0238516 .119999997 17.5751000 .127499998 17.07955187443 3102 50 .135000005 16.6203733 .142499998 16.1993084 .150000006 15.75106177443 3102 51 .159999996 15.2818325 .170000002 14.8558108 .180000007 14.46999417443 3102 52 .189999998 14.0996627 .200000003 13.7648418 .209999993 13.45402097443 3102 53 .219999999 13.1616894 .230000004 12.8935802 .239999995 12.57700827443 3102 54 .254999995 12.2297639 .270000011 11.9645082 .280000001 11.66692567443 3102 55 .300000012 11.3057050 .319999993 10.9795359 .340000004 10.68698387443 3102 56 .360000014 10.4145184 .379999995 10.1634427 .400000006 9.902794257443 3102 57 .425000012 9.63866553 .449999988 9.39853042 .474999994 9.175161707443 3102 58 .500000000 8.97023658 .524999976 8.78189600 .550000012 8.603760717443 3102 59 .574999988 8.44231333 .600000024 8.27353196 .629999995 8.107680887443 3102 60 .660000026 7.94505970 .689999998 7.80132218 .720000029 7.639667707443 3102 61 .759999990 7.47024298 .800000012 7.31302988 .839999974 7.169505637443 3102 62 .879999995 7.04137512 .920000017 6.91613741 .959999979 6.800998867443 3102 63 1.00000000 6.68266043 1.04999995 6.55552625 1.10000002 6.444120707443 3102 64 1.14999998 6.33912340 1.20000005 6.21961754 1.27499998 6.083371227443 3102 65 1.35000002 5.96232397 1.42499995 5.85656846 1.50000000 5.739618607443 3102 66 1.60000002 5.62159377 1.70000005 5.51970966 1.79999995 5.424408957443 3102 67 1.89999998 5.34155762 2.00000000 5.26976356 2.09999990 5.202706237443 3102 68 2.20000005 5.14198715 2.29999995 5.08795501 2.40000010 5.030501777443 3102 69 2.54999995 4.96748721 2.70000005 4.92518185 2.79999995 4.880380877443 3102 70 3.00000000 4.83453936 3.20000005 4.79788239 3.40000010 4.775360897443 3102 71 3.59999990 4.75819048 3.79999995 4.75230086 4.00000000 4.758569807443 3102 72 4.25000000 4.77106693 4.50000000 4.79690169 4.75000000 4.833598287443 3102 73 5.00000000 4.87885727 5.25000000 4.93616534 5.50000000 5.000960757443 3102 74 5.75000000 5.07713265 6.00000000 5.17233026 6.30000019 5.288142407443 3102 75 6.59999990 5.41949101 6.90000010 5.56814022 7.19999981 5.762945927443 3102 76 7.59999990 6.01650068 8.00000000 6.30471484 8.39999962 6.635955477443 3102 77 8.80000019 7.01430912 9.19999981 7.44727175 9.60000038 7.942231267443 3102 78 10.0000000 8.58836367 10.5000000 9.44261389 11.0000000 10.47294877443 3102 79 11.5000000 11.7309584 12.0000000 13.7561740 12.7500000 17.08032167443 3102 80 13.5000000 22.0166931 14.2500000 29.8026254 15.0000000 46.73659427443 3102 81 16.0000000 94.5260670 17.0000000 309.787436 18.0000000 5638.537747443 3102 82 19.0000000 1815.84329 20.0000000 172.939121 21.0000000 62.58393787443 3102 83 22.0000000 31.8812393 23.0000000 19.1580069 24.0000000 11.70253047443 3102 84 25.5000000 7.20800365 27.0000000 5.13275445 28.0000000 3.673188097443 3102 85 30.0000000 2.48695996 32.0000000 1.78528721 34.0000000 1.338528877443 3102 86 36.0000000 1.03821748 38.0000000 .826651283 40.0000000 .6575238467443 3102 87 42.5000000 .522356097 45.0000000 .424603859 47.5000000 .3518773737443 3102 88 50.0000000 .296217390 52.5000000 .252893991 55.0000000 .2185147627443 3102 89 57.5000000 .190829850 60.0000000 .166402823 63.0000000 .1447788107443 3102 90 66.0000000 .127378998 69.0000000 .113118320 72.0000000 .0997255287443 3102 91 76.0000000 .087403753 80.0000000 .077607612 84.0000000 .0697291627443 3102 92 88.0000000 .063342793 92.0000000 .058157718 96.0000000 .0539244977443 3102 93 100.000000 .050163493 105.000000 .046910129 110.000000 .0445344567443 3102 94 115.000000 .042943336 120.000000 .042015733 127.500000 .0425060007443 3102 95 135.000000 .045267428 142.500000 .052012399 150.000000 .0756897997443 3102 96 160.000000 1.13051119 170.000000 45.0652879 180.000000 .1460822507443 3102 97 190.000000 .227195614 200.000000 .749907469 210.000000 99.88707977443 3102 98 220.000000 3.00188776 230.000000 .332455456 240.000000 .1153864727443 3102 99 255.000000 .057977480 270.000000 .064325403 280.000000 13.50414727443 3102 100 300.000000 .024217343 320.000000 .015826335 340.000000 .0144179717443 3102 101 360.000000 .017463390 380.000000 .064570917 400.000000 13.76547697443 3102 102 425.000000 .021313924 450.000000 .019419068 475.000000 .0383031917443 3102 103 500.000000 9.05364172 525.000000 17.2833535 550.000000 .1488100797443 3102 104 575.000000 .043781393 600.000000 .054960947 630.000000 .5885251037443 3102 105 660.000000 15.0661554 690.000000 12.3544576 720.000000 5.594548917443 3102 106 760.000000 .062366218 800.000000 2.14807197 840.000000 .0196663417443 3102 107 880.000000 .023557758 920.000000 .421672481 960.000000 6.765279557443 3102 108 1000.00000 .034585797 1050.00000 4.19808030 1100.00000 3.956369187443 3102 109 1150.00000 3.55440781 1200.00000 .026371997 1275.00000 .0050298987443 3102 110 1350.00000 1.29622275 1425.00000 1.37966813 1500.00000 .9379094897443 3102 111 1600.00000 .126739054 1700.00000 .415754097 1800.00000 .3026193427443 3102 112 1900.00000 1.15273664 2000.00000 1.38820260 2100.00000 .8179907837443 3102 113 2200.00000 .002428240 2300.00000 1.73205297 2400.00000 .7663873497443 3102 114 2550.00000 1.01713316 2700.00000 .633130904 2800.00000 .5180346507443 3102 115 3000.00000 .712914561 3200.00000 .348340976 3400.00000 .6314748127443 3102 116 3600.00000 .594395107 3800.00000 .397666279 4000.00000 .5559907107443 3102 117 4250.00000 .418785330 4500.00000 .433050370 4750.00000 .4190817187443 3102 118 5000.00000 .339884390 5250.00000 .557688526 5500.00000 .4178038907443 3102 119 5750.00000 .388360536 6000.00000 .487815811 6300.00000 .2398880597443 3102 120 6600.00000 .333947252 6900.00000 .464825559 7200.00000 .3477786887443 3102 121 7600.00000 .279407342 8000.00000 .305194564 8400.00000 .2570787697443 3102 122 8800.00000 .316592496 9200.00000 .309418450 9600.00000 .3027768907443 3102 123 10000.0000 .295824112 10500.0000 .288553287 11000.0000 .2817527747443 3102 124 11500.0000 .275422571 12000.0000 .268078367 12750.0000 .2599284077443 3102 125 13500.0000 .252498730 14250.0000 .245686372 15000.0000 .2383856357443 3102 126 16000.0000 .230803327 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139 115000.000 .113007176 120000.000 .102570274 127500.000 .0892110637443 3102 140 135000.000 .079061113 142500.000 .072500254 150000.000 .0675400557443 3102 141 160000.000 .063961465 170000.000 .061367953 180000.000 .0593426637443 3102 142 190000.000 .057674540 200000.000 .056266977 210000.000 .0550179637443 3102 143 220000.000 .053908647 230000.000 .052961052 240000.000 .0519025927443 3102 144 255000.000 .050815358 270000.000 .050028669 280000.000 .0492299807443 3102 145 300000.000 .048361898 320000.000 .047615505 340000.000 .0470352597443 3102 146 360000.000 .046520165 380000.000 .046145041 400000.000 .0458223657443 3102 147 425000.000 .045565832 450000.000 .045400240 475000.000 .0452981907443 3102 148 500000.000 .045310196 525000.000 .045394934 550000.000 .0455146267443 3102 149 575000.000 .045721924 600000.000 .045958155 630000.000 .0462562297443 3102 150 660000.000 .046565762 690000.000 .046875296 720000.000 .0471461887443 3102 151 760000.000 .047171945 800000.000 .046822294 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4.12597- 3 4.07952- 3 4.04296- 3 4.61531- 3 5.96228- 3744333102 65 4.07182- 3 3.04717- 3 2.72009- 3 2.85865- 3 2.82259- 3 2.84474- 3744333102 66 2.72097- 3 3.56800- 3 4.52475- 3 5.12145- 3 5.62007- 3 5.79392- 3744333102 67 5.76950- 3 5.76004- 3 5.72564- 3 5.73317- 3 6.46176- 3 8.39467- 3744333102 68 5.20771- 3 3.71857- 3 3.46976- 3 3.46460- 3 3.50192- 3 3.41207- 3744333102 69 4.46116- 3 5.64115- 3 6.44819- 3 7.01240- 3 7.20858- 3 7.22485- 3744333102 70 7.22639- 3 7.14667- 3 7.15233- 3 8.09342- 3 1.04846- 2 4.89023- 3744333102 71 3.75588- 3 3.59533- 3 3.56163- 3 3.47878- 3 4.58232- 3 5.68368- 3744333102 72 6.55293- 3 7.26525- 3 7.33644- 3 7.29426- 3 7.39179- 3 7.34179- 3744333102 73 7.25344- 3 8.27546- 3 1.07087- 2 4.83225- 3 3.76382- 3 3.75857- 3744333102 74 3.69770- 3 4.82828- 3 6.14842- 3 6.95657- 3 7.58153- 3 7.84944- 3744333102 75 7.84228- 3 7.80493- 3 7.74221- 3 7.77826- 3 8.75400- 3 1.13701- 2744333102 76 5.08337- 3 3.78225- 3 3.58894- 3 4.74728- 3 6.10842- 3 6.75552- 3744333102 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2.35919- 2744333102 90 2.69987- 2 2.38452- 2 1.67894- 2 1.47043- 2 1.94187- 2 2.35998- 2744333102 91 2.63049- 2 2.09755- 2 1.50714- 2 1.83217- 2 2.41037- 2 2.35921- 2744333102 92 1.97452- 2 1.67774- 2 2.36775- 2 2.31409- 2 2.15559- 2 2.28705- 2744333102 93 3.01605- 2 2.94807- 2 4.61913- 2 744333102 94 744333 099999 7443 0 0 0 0 0 0 0 7.91970E+4 1.95274E+2 0 0 34 107925 1451 1 0.0 0.0 0 0 0 67925 1451 2 1.00000E+0 3.00000E+7 0 0 10 20027925 1451 3 3.00000E+2 0.0 1 0 202 67925 1451 4 79-Au-197 LANL/IRK EVAL-JAN84 P.G.YOUNG,VONACH ET AL 7925 1451 5 DIST-Feb2004 7925 1451 6 ----IRDF-2002 MATERIAL 7925 7925 1451 7 -----INCIDENT NEUTRON DATA 7925 1451 8 ------ENDF-6 FORMAT 7925 1451 9 *****************************************************************7925 1451 10 79-AU-197 LANL EVAL-JAN84 P.G.YOUNG 7925 1451 11 LA-10069-PR DIST-SEP91 REV1-JUL91 19930129 7925 1451 12 ----ENDF/B-VI MATERIAL 7925 REVISION 1 7925 1451 13 *****************************************************************7925 1451 14 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************7925 1451 15 DOSIMETRY 7925 1451 16 ******************************************************************7925 1451 17 7925 1451 18 MOD1 OF ENDF/B-VI 7925 1451 19 7925 1451 20 The following revisions were made for MOD1 of ENDF/B-VI: 7925 1451 21 7925 1451 22 1. MF=1,MT=451 - Comments were added regarding estimated 7925 1451 23 (expanded) covariance for the Standards Cross Sections. 7925 1451 24 2. MF=3,MT=102 - Q-value corrected. 7925 1451 25 7925 1451 26 *****************************************************************7925 1451 27 ++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++7925 1451 28 MF/MT 2151, MF/MT 3 16 AND MF/MT 33 16 FROM FOLLOWING 7925 1451 29 EVALUATION 7925 1451 30 79-AU-197 IRK-VIENNA EVAL-APR90 7925 1451 31 DIST-JUN90 7925 1451 32 IRK-EVAL.NLIB 25 7925 7925 1451 33 ++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++++7925 1451 34 *************** SUMMARY OF ENDF/B-VI EVALUATION ******************7925 1451 35 7925 1451 36 A new evaluation of all neutron and gamma-ray data above the 7925 1451 37 resonance region is joined with the endf/b-v resolved resonance 7925 1451 38 region evaluation and with the Version VI standard cross section 7925 1451 39 for the (n,gamma) reaction below a neutron energy of 2.5 MeV. 7925 1451 40 7925 1451 41 *************** GENERAL DESCRIPTION ******************************7925 1451 42 7925 1451 43 P.G.Young and E.D.Arthur 7925 1451 44 7925 1451 45 the new evaluation for files 3,4,5,12,13,14,15 is based on 7925 1451 46 statistical theory, hauser-feshbach, preequilibrium calculations 7925 1451 47 with the comnuc and gnash codes (ref1,2). deformed optical poten-7925 1451 48 tial of delaroche and ecis coupled-channel code were used to cal-7925 1451 49 culate neutron transmission coefficients and total and elastic 7925 1451 50 elastic cross sections (ref3,4). gamma-ray strength functions 7925 1451 51 were obtained by fitting morgan n,xg data (ref5) at 0.4 and 6.5 7925 1451 52 mev. calculated results were used for all major reactions except 7925 1451 53 total cross section. for total, the theoretical cross section 7925 1451 54 was used as prior in covariance analysis of experimental data 7925 1451 55 using glucs code (ref6). more details on experimental data used 7925 1451 56 are given below and in main reference for evaluation (ref 7). 7925 1451 57 7925 1451 58 **************************************************************** 7925 1451 59 7925 1451 60 STANDARDS COVARIANCES 7925 1451 61 7925 1451 62 Phase 1 reviewers of the ENDF/B-VI standards cross sections have 7925 1451 63 expressed the concern that the uncertainties resulting from the 7925 1451 64 combination of R-matrix and simultaneous evaluations might have 7925 1451 65 led to uncertainties that are too small. As a result, the 7925 1451 66 Standards Subcommittee produced (at the May, 1990 CSEWG meeting) 7925 1451 67 a set of expanded covariance estimates for the standard cross 7925 1451 68 section reactions. These uncertainties are estimates such that 7925 1451 69 if a modern day experiment were performed on a given standard 7925 1451 70 cross section using the best techniques, approximately 2/3 of 7925 1451 71 the results should fall within these expanded uncertainties. The 7925 1451 72 expanded uncertainties for the Au-197(n,gamma) cross section are 7925 1451 73 given in the following table and are compared to values from the 7925 1451 74 combined output of the standards covariance analysis: 7925 1451 75 7925 1451 76 Energy Range Estimated Uncertainty Combined Analysis 7925 1451 77 (keV) (percent) (percent) 7925 1451 78 7925 1451 79 2.53E-05 0.14 0.14 7925 1451 80 200 - 500 3.0 1.31 7925 1451 81 500 - 1000 3.5 2.1 7925 1451 82 1000 - 2500 4.5 2.0 7925 1451 83 7925 1451 84 7925 1451 85 *************** mf=2 resonance parameters ************************7925 1451 86 7925 1451 87 mt=151 resolved resonance parameters given from 1.0e-05ev 7925 1451 88 to 2 kev based on ref8 and references therein 7925 1451 89 and a bound level. some of the reson. spin assignments 7925 1451 90 from ref9. from 2 to 4.827 kev the parameters are based7925 1451 91 on macklin et al and hoffman et al normalized data. 7925 1451 92 see refs 10 and 11. 7925 1451 93 thermal cross sections are as follows: 7925 1451 94 capture = 98.71 b 7925 1451 95 scattering = 6.84 b 7925 1451 96 total = 105.55 b 7925 1451 97 the absorption resonance integral is 1559 b 7925 1451 98 7925 1451 99 *************** mf=3 smooth neutron cross sections ***************7925 1451 100 7925 1451 101 mt= 1 total cross section. based on glucs covariance analysis 7925 1451 102 using deformed optical model calculation as the prior and 7925 1451 103 experimental data from refs 12-22, 29 for fitting. 7925 1451 104 mt= 2 elastic cross section. difference of mt=1 and sum of 7925 1451 105 all nonelastic cross sections. closely approximates theore- 7925 1451 106 tical results. 7925 1451 107 mt=102 (n,gamma) cross section. below 2.5 MeV, adopted the 7925 1451 108 ENDF/B-VI standard cross section (Ref.30,31) down to the 7925 1451 109 resonance region. At higher energies, the theoretical cal- 7925 1451 110 culations were adjusted to agree with experimental data. A 7925 1451 111 semi-direct component normalized to an average of experimental7925 1451 112 data at 14 MeV was included above En = 6 MeV. 7925 1451 113 at higher energies, use theoretical calculations, which agree 7925 1451 114 reasonably with available exp. data. above 5 mev, calculation7925 1451 115 includes semi-direct component normalized to average of 7925 1451 116 14 mev data. 7925 1451 117 7925 1451 118 *************** mf=33 neutron cross section covariances **********7925 1451 119 7925 1451 120 mt= 1 total cross section covariance from glucs analysis. 7925 1451 121 7925 1451 122 *************** references ***************************************7925 1451 123 7925 1451 124 1. c.l.dunford. ai-aec-12931(1970) 7925 1451 125 2. p.g.young, e.d.arthur, la-6947 (1977). 7925 1451 126 3. j.p.delaroche, harwell conference (1978)p.366. 7925 1451 127 4. j.raynal, iaea smr-9/8 (1970). 7925 1451 128 5. g.l.morgan, e.newman, ornl-tm-4973 (1975). 7925 1451 129 6. d.m.hetrick, c.y.fu, ornl/tm-7341 (1980). 7925 1451 130 7. p.g.young, e.d.arthur, in la-10069-pr (1984)p.12. 7925 1451 131 8. s.f.mughabghab and d.i.garber bnl-325,3rd edn,vol i(1973). 7925 1451 132 9. a.lottin and a.jain conf on nuclear structure study with 7925 1451 133 neutrons,budapest,1972 p34 and private communication. 7925 1451 134 10. r.macklin et al. phys. rev/c 11,1270(1975) and private 7925 1451 135 communication. 7925 1451 136 11. m.m. hoffman et al. 71knoxville conf., 2, 868(1971) 7925 1451 137 12. w.poenitz et al., nuc.sci.eng. 78, 333(1981). 7925 1451 138 13. d.g.foster jr., d.glasgow, phys.rev. c3, 576(1971). 7925 1451 139 14. k.k.seth,phys.letters,16,306(1965). 7925 1451 140 15. s.c.snowdon, phys.rev. 90, 615(1953). 7925 1451 141 16. j.f.whalen,anl-7210,16(1966). 7925 1451 142 17. n.nereson, phys.rev. 94, 1678(1954). 7925 1451 143 18. a.bratenahl et al., phys.rev. 110, 927(1958). 7925 1451 144 19. j.p.conner,phys.rev.109,1268(1958). 7925 1451 145 20. j.h.coon,phys.rev.88,562(1952). 7925 1451 146 21. j.m.peterson,phys.rev.120,521(1960). 7925 1451 147 22. e.g.bilpuch,private communication(1959). 7925 1451 148 23. j.frehaut et al, proc. 10-50 mev conf, bnl-ncs-51245 (1980) 7925 1451 149 page 399. 7925 1451 150 24. l.r.veeser et al, phys.rev. c16, 1792(1977). 7925 1451 151 25. b.p.bayhurst et al, phys.rev. c12, 451(1975). 7925 1451 152 26. r.j.prestwood and b.p.bayhurst,phys.rev.121,1438(1961). 7925 1451 153 27. c.kalbach and f.mann, bnl-ncs-5/245,p.689 (1980). 7925 1451 154 28. v.j.orphan et al, ga-10248 (1970). 7925 1451 155 29. d.c.larson, proc. 10-50 mev conf, bnl-ncs-51245 (1980) p.277.7925 1451 156 30. A.Carlson et al., Nuc.Data for Basic & Applied Science, 7925 1451 157 Santa Fe, NM (1985) p.1429. 7925 1451 158 31. W.Poenitz, ANL-West, personnal communication (1989). 7925 1451 159 7925 1451 160 ******************************************************************7925 1451 161 7925 1451 162 7925 1451 163 7925 1451 164 7925 1451 165 ***************** PROGRAM FIXUP (VERSION 86-2) ******************7925 1451 166 *RECONSTRUCTED MT NUMBERS 7925 1451 167 4 =+( 51, 91) 7925 1451 168 103 =+(700,718) 7925 1451 169 104 =+(720,738) 7925 1451 170 105 =+(740,758) 7925 1451 171 106 =+(760,778) 7925 1451 172 107 =+(780,798) 7925 1451 173 101 =+(102,114) 7925 1451 174 27 =+( 18, 18)+(101,101) 7925 1451 175 3 =+( 4, 4)+( 6, 9)+( 16, 17)+( 22, 37) 7925 1451 176 19 =+( 18, 18)-( 20, 21)-( 38, 38) 7925 1451 177 1 =+( 2, 3) 7925 1451 178 ***************** Program LINEAR (VERSION 2002-1) ***************7925 1451 179 For All Data Greater than 1.0000E-10 barns in Absolute Value 7925 1451 180 Data Linearized to Within an Accuracy of .100000000 per-cent 7925 1451 181 ***************** Program RECENT (VERSION 2002-1) ***************7925 1451 182 for All Data Greater than 1.0000E-10 barns in Absolute Value 7925 1451 183 Data Linearized to within an Accuracy of .100000000 per-cent 7925 1451 184 ***************** Program SIGMA1 (VERSION 2002-1) ***************7925 1451 185 Data Doppler Broadened to 300.000000 Kelvin 7925 1451 186 for All Data Greater than 1.0000E-10 barns in Absolute Value 7925 1451 187 Data Linearized to Within an Accuracy pf .100000000 per-cent 7925 1451 188 ***************** Program FIXUP (Version 2002-1) ****************7925 1451 189 Corrected ZA/AWR in All Sections-----------------------------Yes 7925 1451 190 Corrected Thresholds-----------------------------------------Yes 7925 1451 191 Extended Cross Sections to 20 MeV----------------------------No 7925 1451 192 Allow Cross Section Deletion---------------------------------No 7925 1451 193 Allow Cross Section Reconstruction---------------------------No 7925 1451 194 Make All Cross Sections Non-Negative-------------------------Yes 7925 1451 195 Delete Energies Not in Ascending Order-----------------------Yes 7925 1451 196 Deleted Duplicate Points-------------------------------------Yes 7925 1451 197 Check for Ascending MAT/MF/MT Order--------------------------Yes 7925 1451 198 Check for Legal MF/MT Numbers--------------------------------Yes 7925 1451 199 Allow Creation of Missing Sections---------------------------No 7925 1451 200 Allow Insertion of Energy Points-----------------------------No 7925 1451 201 Create Uniform Energy Grid-----------------------------------No 7925 1451 202 Delete Section if Cross Section =0 at All Energies-----------Yes 7925 1451 203 ***************** Program GROUPIE (VERSION 2002-1) **************7925 1451 204 Unshielded Group Averages Using 640 Groups 7925 1451 205 Weighting Spectrum: Flat (Constant) Spectrum 7925 1451 206 1 451 212 27925 1451 207 2 151 4 07925 1451 208 3 16 43 17925 1451 209 3 102 217 27925 1451 210 33 16 71 17925 1451 211 33 102 26 17925 1451 212 7925 1 099999 7925 0 0 0 7.91970E+4 1.95275E+2 0 0 1 07925 2151 1 7.91970E+4 1.00000E+0 0 0 1 07925 2151 2 1.00000E-5 5.00000E+3 0 0 0 07925 2151 3 1.50000E+0 9.80000E-1 0 0 0 07925 2151 4 7925 2 099999 7925 0 0 0 7.91970E+4 1.95274E+2 0 0 0 07925 3 16 1 -8.08000E+6-8.08000E+6 0 0 1 1207925 3 16 2 120 1 7925 3 16 3 8100000.00 .005579896 8200000.00 .023230150 8300000.00 .0417860347925 3 16 4 8400000.00 .071166000 8500000.00 .104974000 8600000.00 .1404245007925 3 16 5 8700000.00 .179890000 8800000.00 .220456917 8900000.00 .2812400007925 3 16 6 9000000.00 .350293333 9100000.00 .419346667 9200000.00 .4888333337925 3 16 7 9300000.00 .560920000 9400000.00 .633440000 9500000.00 .7059600007925 3 16 8 9600000.00 .778480000 9700000.00 .851136750 9800000.00 .9246140007925 3 16 9 9900000.00 .998228000 10000000.0 1.07184200 10100000.0 1.145456007925 3 16 10 10200000.0 1.21760825 10300000.0 1.28099000 10400000.0 1.342910007925 3 16 11 10500000.0 1.40483000 10600000.0 1.46675000 10700000.0 1.524579007925 3 16 12 10800000.0 1.55786200 10900000.0 1.58705400 11000000.0 1.616246007925 3 16 13 11100000.0 1.64543800 11200000.0 1.67711725 11300000.0 1.723720007925 3 16 14 11400000.0 1.77281000 11500000.0 1.82190000 11600000.0 1.870990007925 3 16 15 11700000.0 1.91616500 11800000.0 1.93785000 11900000.0 1.955620007925 3 16 16 12000000.0 1.97339000 12100000.0 1.99116000 12200000.0 2.008416687925 3 16 17 12300000.0 2.02259345 12400000.0 2.03625690 12500000.0 2.049920347925 3 16 18 12600000.0 2.06358379 12700000.0 2.07724724 12800000.0 2.090910697925 3 16 19 12900000.0 2.10411356 13000000.0 2.10718360 13100000.0 2.106108407925 3 16 20 13200000.0 2.10503320 13300000.0 2.10395800 13400000.0 2.102882807925 3 16 21 13500000.0 2.10180760 13600000.0 2.10313545 13700000.0 2.106866367925 3 16 22 13800000.0 2.11048693 13900000.0 2.11168000 14000000.0 2.111880007925 3 16 23 14100000.0 2.11336363 14200000.0 2.11798500 14300000.0 2.122594637925 3 16 24 14400000.0 2.12380800 14500000.0 2.12363200 14600000.0 2.124854387925 3 16 25 14700000.0 2.12949500 14800000.0 2.13413834 14900000.0 2.135423147925 3 16 26 15000000.0 2.13533400 15100000.0 2.13524486 15200000.0 2.135155717925 3 16 27 15300000.0 2.13506657 15400000.0 2.13497743 15500000.0 2.134888297925 3 16 28 15600000.0 2.13479914 15700000.0 2.13272231 15800000.0 2.118719337925 3 16 29 15900000.0 2.10272867 16000000.0 2.08673800 16100000.0 2.070747337925 3 16 30 16200000.0 2.05475667 16300000.0 2.03876600 16400000.0 2.022775337925 3 16 31 16500000.0 2.00678467 16600000.0 1.99079400 16700000.0 1.974803337925 3 16 32 16800000.0 1.95881267 16900000.0 1.94282200 17000000.0 1.926831337925 3 16 33 17100000.0 1.91084067 17200000.0 1.89294833 17300000.0 1.863646007925 3 16 34 17400000.0 1.83244200 17500000.0 1.80123800 17600000.0 1.770034007925 3 16 35 17700000.0 1.73883000 17800000.0 1.70762600 17900000.0 1.676422007925 3 16 36 18000000.0 1.64521800 18100000.0 1.61401400 18200000.0 1.582810007925 3 16 37 18300000.0 1.55160600 18400000.0 1.52040200 18500000.0 1.483501767925 3 16 38 18600000.0 1.44090529 18700000.0 1.39830882 18800000.0 1.355712357925 3 16 39 18900000.0 1.31311588 19000000.0 1.27051941 19100000.0 1.227922947925 3 16 40 19200000.0 1.18532647 19300000.0 1.14579943 19400000.0 1.124689007925 3 16 41 19500000.0 1.10664800 19600000.0 1.08860700 19700000.0 1.070566007925 3 16 42 19800000.0 1.05252500 19900000.0 1.03448400 20000000.0 0.0 7925 3 16 43 7925 3 099999 7.91970E+4 1.95274E+2 0 0 0 07925 3102 1 6.51238E+6 6.51238E+6 0 0 1 6417925 3102 2 641 1 7925 3102 3 .000100000 1537.29252 .000105000 1501.44899 .000110000 1467.397307925 3102 4 .000115000 1436.01995 .000120000 1399.33870 .000127500 1358.641447925 3102 5 .000135000 1321.45967 .000142500 1287.29063 .000150000 1250.310417925 3102 6 .000160000 1211.78680 .000170000 1176.73521 .000180000 1144.560717925 3102 7 .000190000 1114.62516 .000200000 1087.29347 .000210000 1061.579027925 3102 8 .000220000 1037.69457 .000230000 1015.50287 .000240000 989.4543447925 3102 9 .000255000 960.831039 .000270000 938.744134 .000280000 914.1573887925 3102 10 .000300000 884.173150 .000320000 857.023494 .000340000 832.1588477925 3102 11 .000360000 809.257102 .000380000 788.350966 .000400000 766.5551987925 3102 12 .000425000 744.251380 .000450000 723.872005 .000475000 705.1891857925 3102 13 .000500000 687.619941 .000525000 671.562452 .000550000 656.3934707925 3102 14 .000575000 642.285605 .000600000 627.826066 .000630000 612.9510547925 3102 15 .000660000 599.340433 .000690000 586.284077 .000720000 572.3914287925 3102 16 .000760000 557.495391 .000800000 543.664916 .000840000 530.9903397925 3102 17 .000880000 518.943943 .000920000 507.900667 .000960000 497.3246297925 3102 18 .001000000 486.362115 .001050000 474.918645 .001100000 464.1913267925 3102 19 .001150000 454.307055 .001200000 442.647469 .001275000 429.8334017925 3102 20 .001350000 418.118588 .001425000 407.186240 .001500000 395.5727517925 3102 21 .001600000 383.430809 .001700000 372.347369 .001800000 362.0972177925 3102 22 .001900000 352.734017 .002000000 344.052418 .002100000 335.9056627925 3102 23 .002200000 328.445141 .002300000 321.311073 .002400000 313.1706647925 3102 24 .002550000 304.124389 .002700000 297.047972 .002800000 289.3582737925 3102 25 .003000000 279.892094 .003200000 271.278970 .003400000 263.4050697925 3102 26 .003600000 256.227562 .003800000 249.603436 .004000000 242.7024357925 3102 27 .004250000 235.699851 .004500000 229.309233 .004750000 223.3359257925 3102 28 .005000000 217.852235 .005250000 212.765682 .005500000 207.9548987925 3102 29 .005750000 203.542253 .006000000 198.906549 .006300000 194.2713607925 3102 30 .006600000 189.886414 .006900000 185.822166 .007200000 181.4082817925 3102 31 .007600000 176.693912 .008000000 172.390086 .008400000 168.3883327925 3102 32 .008800000 164.712392 .009200000 161.105934 .009600000 157.8468347925 3102 33 .010000000 154.330462 .010500000 150.747366 .011000000 147.3700377925 3102 34 .011500000 144.221144 .012000000 140.584965 .012750000 136.5312257925 3102 35 .013500000 132.824467 .014250000 129.435415 .015000000 125.7629757925 3102 36 .016000001 121.935514 .017000001 118.450410 .017999999 115.2633947925 3102 37 .018999999 112.286014 .020000000 109.561077 .021000000 107.0224547925 3102 38 .022000000 104.638524 .023000000 102.465150 .024000000 99.85883977925 3102 39 .025500000 97.0380020 .027000001 94.8246650 .028000001 92.42430877925 3102 40 .029999999 89.4684790 .032000002 86.7734053 .034000002 84.31140787925 3102 41 .035999998 82.0650200 .037999999 79.9947802 .039999999 77.83657157925 3102 42 .042500000 75.6484100 .045000002 73.6514483 .047499999 71.79363757925 3102 43 .050000001 70.0886291 .052499998 68.5162423 .055000000 67.03301467925 3102 44 .057500001 65.6958772 .059999999 64.2644203 .063000001 62.81128337925 3102 45 .066000000 61.4578421 .068999998 60.2146434 .071999997 58.83941687925 3102 46 .075999998 57.4021479 .079999998 56.0670142 .083999999 54.82132787925 3102 47 .088000000 53.6810646 .092000000 52.5913686 .096000001 51.59775847925 3102 48 .100000001 50.5297939 .104999997 49.4450140 .109999999 48.42801377925 3102 49 .115000002 47.4819972 .119999997 46.3875128 .127499998 45.16472307925 3102 50 .135000005 44.0503141 .142499998 43.0362057 .150000006 41.93990547925 3102 51 .159999996 40.8057609 .170000002 39.7944610 .180000007 38.83650607925 3102 52 .189999998 37.9911792 .200000003 37.1728634 .209999993 36.44175077925 3102 53 .219999999 35.7867340 .230000004 35.1426385 .239999995 34.42348767925 3102 54 .254999995 33.6154139 .270000011 33.0136267 .280000001 32.33737407925 3102 55 .300000012 31.5248164 .319999993 30.8047948 .340000004 30.16142567925 3102 56 .360000014 29.5868954 .379999995 29.0404626 .400000006 28.51089197925 3102 57 .425000012 27.9618870 .449999988 27.4882584 .474999994 27.04839567925 3102 58 .500000000 26.6687770 .524999976 26.3039381 .550000012 25.99200867925 3102 59 .574999988 25.7259568 .600000024 25.4414972 .629999995 25.17634067925 3102 60 .660000026 24.9381122 .689999998 24.7209218 .720000029 24.52089797925 3102 61 .759999990 24.3100987 .800000012 24.1596553 .839999974 24.02185247925 3102 62 .879999995 23.9374044 .920000017 23.8642712 .959999979 23.81519297925 3102 63 1.00000000 23.8008509 1.04999995 23.8273729 1.10000002 23.88641777925 3102 64 1.14999998 23.9797452 1.20000005 24.1184559 1.27499998 24.36020907925 3102 65 1.35000002 24.6743302 1.42499995 25.0446067 1.50000000 25.56475497925 3102 66 1.60000002 26.2693686 1.70000005 27.0828354 1.79999995 28.03658207925 3102 67 1.89999998 29.1320605 2.00000000 30.3825863 2.09999990 31.79511597925 3102 68 2.20000005 33.4102728 2.29999995 35.2433462 2.40000010 37.91256777925 3102 69 2.54999995 41.7098948 2.70000005 45.5197409 2.79999995 51.18692347925 3102 70 3.00000000 60.9691962 3.20000005 74.5978554 3.40000010 94.41603557925 3102 71 3.59999990 124.824071 3.79999995 174.919844 4.00000000 286.0010437925 3102 72 4.25000000 617.301275 4.50000000 2459.39283 4.75000000 18409.26697925 3102 73 5.00000000 4282.77397 5.25000000 728.833098 5.50000000 291.2255427925 3102 74 5.75000000 154.998188 6.00000000 91.9087531 6.30000019 58.06983507925 3102 75 6.59999990 39.7909171 6.90000010 28.8500384 7.19999981 20.93280267925 3102 76 7.59999990 15.1949028 8.00000000 11.4939378 8.39999962 8.979720707925 3102 77 8.80000019 7.19682944 9.19999981 5.88842827 9.60000038 4.902696997925 3102 78 10.0000000 4.06426786 10.5000000 3.35862380 11.0000000 2.822208247925 3102 79 11.5000000 2.40506334 12.0000000 2.00713668 12.7500000 1.646472067925 3102 80 13.5000000 1.37815674 14.2500000 1.17294347 15.0000000 .9909460337925 3102 81 16.0000000 .833272462 17.0000000 .713942300 18.0000000 .6219160237925 3102 82 19.0000000 .549403977 20.0000000 .491655073 21.0000000 .4450035537925 3102 83 22.0000000 .407044594 23.0000000 .375831363 24.0000000 .3445336307925 3102 84 25.5000000 .315411599 27.0000000 .296723395 28.0000000 .2802886707925 3102 85 30.0000000 .265673723 32.0000000 .258550107 34.0000000 .2583786107925 3102 86 36.0000000 .265360233 38.0000000 .280505870 40.0000000 .3109225867925 3102 87 42.5000000 .375780029 45.0000000 2.12704336 47.5000000 .6352437287925 3102 88 50.0000000 .940577311 52.5000000 1.74316513 55.0000000 5.482857417925 3102 89 57.5000000 129.955698 60.0000000 550.969463 63.0000000 3.949048027925 3102 90 66.0000000 1.28229183 69.0000000 .723558548 72.0000000 .8039275997925 3102 91 76.0000000 71.5630182 80.0000000 .912309009 84.0000000 .2433926997925 3102 92 88.0000000 .153249875 92.0000000 .125746947 96.0000000 .1318782317925 3102 93 100.000000 .334418537 105.000000 34.8157915 110.000000 .2085292567925 3102 94 115.000000 .109596811 120.000000 2.28258883 127.500000 .0887210787925 3102 95 135.000000 .230357411 142.500000 12.4158937 150.000000 31.44216507925 3102 96 160.000000 47.2490655 170.000000 .188508247 180.000000 4.571612067925 3102 97 190.000000 22.1637923 200.000000 .773421397 210.000000 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2 0.000000+0 0.000000+0 0 1 8 4792533102 3 1.000000-5 0.000000+0 2.900000+6 5.000000-2 1.000000+7 6.250000-2792533102 4 2.000000+7 0.000000+0 792533102 5 0.000000+0 0.000000+0 1 5 120 15792533102 6 1.000000-5 3.000000-1 2.000000+3 1.000000+4 3.750000+4 9.000000+4792533102 7 1.350000+5 2.050000+5 2.750000+5 4.125000+5 6.250000+5 9.500000+5792533102 8 1.325000+6 1.900000+6 2.900000+6 1.968800-6 1.968800-6 1.197701-7792533102 9 1.277322-7 1.199737-7 1.180243-7 1.210327-7 1.175025-7 1.184462-7792533102 10 1.362774-7 1.098021-7 1.079868-7 1.089638-7 9.914542-8 2.968800-6792533102 11 1.197701-7 1.277322-7 1.199737-7 1.180243-7 1.210327-7 1.175025-7792533102 12 1.184462-7 1.362774-7 1.098021-7 1.079868-7 1.089638-7 9.914542-8792533102 13 1.094560-4 4.577940-5 4.677039-5 3.948904-5 3.768945-5 3.747630-5792533102 14 3.690735-5 3.428047-5 3.338854-5 3.229846-5 3.154064-5 2.637712-5792533102 15 6.056941-5 4.202671-5 3.750888-5 3.574985-5 3.503839-5 3.439512-5792533102 16 3.204558-5 3.144094-5 3.085597-5 2.984776-5 2.493463-5 6.285997-5792533102 17 4.547523-5 4.201196-5 4.049420-5 3.979974-5 3.715363-5 3.659218-5792533102 18 3.509315-5 3.465730-5 2.895877-5 8.021999-5 4.906673-5 4.591913-5792533102 19 4.475850-5 4.261771-5 4.283861-5 4.025727-5 4.041204-5 3.382565-5792533102 20 8.028292-5 5.304576-5 4.689840-5 4.483827-5 4.540301-5 4.299516-5792533102 21 4.362940-5 3.699681-5 9.958098-5 5.534255-5 4.876851-5 5.031469-5792533102 22 4.706249-5 4.804336-5 4.072953-5 9.094943-5 5.454792-5 5.269018-5792533102 23 4.994396-5 5.121352-5 4.343583-5 8.936425-5 5.920198-5 5.242733-5792533102 24 5.357873-5 4.570034-5 1.241536-4 6.756325-5 6.367727-5 5.555147-5792533102 25 1.336240-4 7.281097-5 5.539927-5 1.846612-4 8.162500-5 2.399631-4792533102 26 792533 099999 7925 0 0 0 0 0 0 0 8.01990E+4 1.97259E+2 0 0 34 108034 1451 1 0.0 0.0 0 0 0 68034 1451 2 1.00000E+0 2.00000E+7 0 0 10 20028034 1451 3 3.00000E+2 0.0 1 0 78 58034 1451 4 80-Hg-199 JAERI EVAL-MAR90 K.SAKURAI 8034 1451 5 DIST-Feb2004 8034 1451 6 ----IRDF-2002 MATERIAL 8034 8034 1451 7 -----INCIDENT NEUTRON DATA 8034 1451 8 ------ENDF-6 FORMAT 8034 1451 9 *************************************************************** 8034 1451 10 80-HG-199 JAERI EVAL-MAR90 K.SAKURAI 8034 1451 11 DIST-JUL91 8034 1451 12 ----JENDL/D-99 MATERIAL 8034 8034 1451 13 *************************************************************** 8034 1451 14 HISTORY 8034 1451 15 90-03 EVALUATION WAS MADE BY K. SAKURA(JAERI). 8034 1451 16 90-10 COMPILED TO JENDL DOSIMETRY FILE VERSION 1. 8034 1451 17 91-07 DESCRIPTIVE DATA WERE ADDED. 8034 1451 18 8034 1451 19 ===== POINT-WISE DATA FILE ===== 8034 1451 20 8034 1451 21 8034 1451 22 HG-199 (N,N')HG-199M (HALF-LIFE = 42.6 M) 8034 1451 23 8034 1451 24 MF=1 GENERAL INFORMATION 8034 1451 25 MT=451 DESCRIPTIVE DATA AND DICTIONARY 8034 1451 26 8034 1451 27 MF=2 RESONANCE PARAMETERS 8034 1451 28 MT=151 PARAMETERS 8034 1451 29 ONLY SPIN AND SCATTERING RADIUS ARE GIVEN. 8034 1451 30 8034 1451 31 MF=3 NEUTRON CROSS SECTIONS 8034 1451 32 MT=51 HG-199(N,N')HG-199M CROSS SECTION 8034 1451 33 ISOMERIC STATE = 0.531 MEV 13/2+, HALF-LIFE = 42.6 M 8034 1451 34 EVALUATION WAS BASED ON THE EXPERIMENTAL DATA REPORTED IN 8034 1451 35 REF./1/. 8034 1451 36 8034 1451 37 MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS 8034 1451 38 MT=51 8034 1451 39 ESTIMATED FROM THE EXPERIMENTAL DATA. 8034 1451 40 8034 1451 41 REFERENCE 8034 1451 42 1) SAKURAI K., ET AL.: J. NUCL. SCI. TECHNOL., 19, 775 (1982). 8034 1451 43 8034 1451 44 *************************************************************** 8034 1451 45 ----- MF=3 MT=51 ----- 8034 1451 46 For the IRDF-2002 file this reaction was converted at 8034 1451 47 IAEA/NDS. 8034 1451 48 The reaction MF/MT=3/51 was converted to MF/MT=10/ 4 8034 1451 49 The corresponding co-variance files were also converted 8034 1451 50 The reaction MF/MT=33/51 was converted to MF/MT=40/ 4 8034 1451 51 *************************************************************** 8034 1451 52 The threshold energy was recalculated from 5.25000+5 to 8034 1451 53 5.33691+5 to satisfy code FIZCON criteria 8034 1451 54 8034 1451 55 8034 1451 56 *************************************************************** 8034 1451 57 ***************** Program LINEAR (VERSION 2002-1) ***************8034 1451 58 For All Data Greater than 1.0000E-10 barns in Absolute Value 8034 1451 59 Data Linearized to Within an Accuracy of .100000000 per-cent 8034 1451 60 ***************** Program SIGMA1 (VERSION 2002-1) ***************8034 1451 61 Data Doppler Broadened to 300.000000 Kelvin 8034 1451 62 for All Data Greater than 1.0000E-10 barns in Absolute Value 8034 1451 63 Data Linearized to Within an Accuracy pf .100000000 per-cent 8034 1451 64 ***************** Program FIXUP (Version 2002-1) ****************8034 1451 65 Corrected ZA/AWR in All Sections-----------------------------Yes 8034 1451 66 Corrected Thresholds-----------------------------------------Yes 8034 1451 67 Extended Cross Sections to 20 MeV----------------------------No 8034 1451 68 Allow Cross Section Deletion---------------------------------No 8034 1451 69 Allow Cross Section Reconstruction---------------------------No 8034 1451 70 Make All Cross Sections Non-Negative-------------------------Yes 8034 1451 71 Delete Energies Not in Ascending Order-----------------------Yes 8034 1451 72 Deleted Duplicate Points-------------------------------------Yes 8034 1451 73 Check for Ascending MAT/MF/MT Order--------------------------Yes 8034 1451 74 Check for Legal MF/MT Numbers--------------------------------Yes 8034 1451 75 Allow Creation of Missing Sections---------------------------No 8034 1451 76 Allow Insertion of Energy Points-----------------------------No 8034 1451 77 Create Uniform Energy Grid-----------------------------------No 8034 1451 78 Delete Section if Cross Section =0 at All Energies-----------Yes 8034 1451 79 ***************** Program GROUPIE (VERSION 2002-1) **************8034 1451 80 Unshielded Group Averages Using 640 Groups 8034 1451 81 Weighting Spectrum: Flat (Constant) Spectrum 8034 1451 82 1 451 87 08034 1451 83 2 151 4 08034 1451 84 8 4 2 08034 1451 85 10 4 72 08034 1451 86 40 4 17 08034 1451 87 8034 1 099999 8034 0 0 0 8.01990E+4 1.97259E+2 0 0 1 08034 2151 1 8.01990E+4 1.00000E+0 0 0 1 08034 2151 2 0.0 0.0 0 0 0 08034 2151 3 5.00000E-1 1.03000E+0 0 0 0 08034 2151 4 8034 2 099999 8034 0 0 0 8.01990E+4 1.97259E+2 0 0 1 18034 8 4 1 8.01990E+4 0.000000+0 10 0 0 08034 8 4 2 8034 8 099999 8034 0 0 0 8.01990E+4 1.97259E+2 0 0 1 0803410 4 1 0.0 -5.30999E+5 80199 0 1 205803410 4 2 205 1 803410 4 3 525000.000 .000774677 550000.000 .004195311 575000.000 .007835936803410 4 4 600000.000 .011840624 630000.000 .016209374 660000.000 .020578124803410 4 5 690000.000 .024946875 720000.000 .030024219 760000.000 .035400000803410 4 6 800000.000 .040600000 840000.000 .045800000 880000.000 .051000000803410 4 7 920000.000 .056200000 960000.000 .061400000 1000000.00 .071400000803410 4 8 1100000.00 .086200000 1200000.00 .102100000 1300000.00 .124600000803410 4 9 1400000.00 .148200000 1500000.00 .174000000 1600000.00 .202000000803410 4 10 1700000.00 .229000000 1800000.00 .250000000 1900000.00 .270000000803410 4 11 2000000.00 .290000000 2100000.00 .310000000 2200000.00 .329250000803410 4 12 2300000.00 .344000000 2400000.00 .358000000 2500000.00 .375000000803410 4 13 2600000.00 .395000000 2700000.00 .414250000 2800000.00 .429000000803410 4 14 2900000.00 .443000000 3000000.00 .457000000 3100000.00 .471000000803410 4 15 3200000.00 .484000000 3300000.00 .491000000 3400000.00 .497000000803410 4 16 3500000.00 .506000000 3600000.00 .518000000 3700000.00 .529500000803410 4 17 3800000.00 .538000000 3900000.00 .546000000 4000000.00 .554000000803410 4 18 4100000.00 .562000000 4200000.00 .569500000 4300000.00 .574000000803410 4 19 4400000.00 .578000000 4500000.00 .582000000 4600000.00 .586000000803410 4 20 4700000.00 .590000000 4800000.00 .594000000 4900000.00 .598000000803410 4 21 5000000.00 .600400000 5100000.00 .601200000 5200000.00 .602300000803410 4 22 5300000.00 .605200000 5400000.00 .608400000 5500000.00 .612000000803410 4 23 5600000.00 .616000000 5700000.00 .619550000 5800000.00 .620400000803410 4 24 5900000.00 .620800000 6000000.00 .622900000 6100000.00 .626700000803410 4 25 6200000.00 .630500000 6300000.00 .634300000 6400000.00 .638100000803410 4 26 6500000.00 .640000000 6600000.00 .640000000 6700000.00 .640000000803410 4 27 6800000.00 .640000000 6900000.00 .640000000 7000000.00 .640000000803410 4 28 7100000.00 .640000000 7200000.00 .640000000 7300000.00 .640000000803410 4 29 7400000.00 .640000000 7500000.00 .640000000 7600000.00 .640000000803410 4 30 7700000.00 .640000000 7800000.00 .640000000 7900000.00 .640000000803410 4 31 8000000.00 .637000000 8100000.00 .631000000 8200000.00 .625625000803410 4 32 8300000.00 .624000000 8400000.00 .623000000 8500000.00 .622000000803410 4 33 8600000.00 .621000000 8700000.00 .619625000 8800000.00 .616000000803410 4 34 8900000.00 .612000000 9000000.00 .606000000 9100000.00 .598000000803410 4 35 9200000.00 .590500000 9300000.00 .586000000 9400000.00 .582000000803410 4 36 9500000.00 .576000000 9600000.00 .568000000 9700000.00 .559500000803410 4 37 9800000.00 .548000000 9900000.00 .536000000 10000000.0 .526000000803410 4 38 10100000.0 .518000000 10200000.0 .510500000 10300000.0 .506000000803410 4 39 10400000.0 .502000000 10500000.0 .494000000 10600000.0 .482000000803410 4 40 10700000.0 .470500000 10800000.0 .462000000 10900000.0 .454000000803410 4 41 11000000.0 .442000000 11100000.0 .426000000 11200000.0 .410500000803410 4 42 11300000.0 .398000000 11400000.0 .386000000 11500000.0 .374000000803410 4 43 11600000.0 .362000000 11700000.0 .349000000 11800000.0 .330000000803410 4 44 11900000.0 .310000000 12000000.0 .296000000 12100000.0 .288000000803410 4 45 12200000.0 .279500000 12300000.0 .268000000 12400000.0 .256000000803410 4 46 12500000.0 .244000000 12600000.0 .232000000 12700000.0 .220500000803410 4 47 12800000.0 .212000000 12900000.0 .204000000 13000000.0 .196000000803410 4 48 13100000.0 .188000000 13200000.0 .180500000 13300000.0 .176000000803410 4 49 13400000.0 .172000000 13500000.0 .168000000 13600000.0 .164000000803410 4 50 13700000.0 .160000000 13800000.0 .156000000 13900000.0 .152000000803410 4 51 14000000.0 .148000000 14100000.0 .144000000 14200000.0 .140000000803410 4 52 14300000.0 .136000000 14400000.0 .132000000 14500000.0 .128000000803410 4 53 14600000.0 .124000000 14700000.0 .120250000 14800000.0 .118000000803410 4 54 14900000.0 .116000000 15000000.0 .114000000 15100000.0 .112000000803410 4 55 15200000.0 .110000000 15300000.0 .108000000 15400000.0 .106000000803410 4 56 15500000.0 .104600000 15600000.0 .103800000 15700000.0 .102950000803410 4 57 15800000.0 .101800000 15900000.0 .100600000 16000000.0 .099600000803410 4 58 16100000.0 .098800000 16200000.0 .098050000 16300000.0 .097600000803410 4 59 16400000.0 .097200000 16500000.0 .096400000 16600000.0 .095200000803410 4 60 16700000.0 .093950000 16800000.0 .092400000 16900000.0 .090800000803410 4 61 17000000.0 .089800000 17100000.0 .089400000 17200000.0 .089000000803410 4 62 17300000.0 .088600000 17400000.0 .088200000 17500000.0 .087800000803410 4 63 17600000.0 .087400000 17700000.0 .087000000 17800000.0 .086600000803410 4 64 17900000.0 .086200000 18000000.0 .085800000 18100000.0 .085400000803410 4 65 18200000.0 .085000000 18300000.0 .084600000 18400000.0 .084200000803410 4 66 18500000.0 .083800000 18600000.0 .083400000 18700000.0 .083000000803410 4 67 18800000.0 .082600000 18900000.0 .082200000 19000000.0 .081800000803410 4 68 19100000.0 .081400000 19200000.0 .081050000 19300000.0 .081000000803410 4 69 19400000.0 .081000000 19500000.0 .080800000 19600000.0 .080400000803410 4 70 19700000.0 .080050000 19800000.0 .080000000 19900000.0 .080000000803410 4 71 20000000.0 0.0 803410 4 72 803410 099999 8034 0 0 0 8.01990E+4 1.97259E+2 0 0 1 0803440 4 1 0.000000+0-5.309990+5 0 0 0 1803440 4 2 1.000000+1 0.000000+0 0 4 0 2803440 4 3 0.000000+0 0.000000+0 0 1 62 31803440 4 4 1.000000-5 0.000000+0 5.250000+5 2.464000-1 7.500000+5 8.640000-2803440 4 5 1.000000+6 1.600000-2 2.000000+6 1.600000-2 2.250000+6 2.200000-2803440 4 6 3.000000+6 2.200000-2 3.500000+6 1.600000-2 4.000000+6 1.600000-2803440 4 7 5.000000+6 1.600000-2 6.000000+6 1.600000-2 7.000000+6 1.600000-2803440 4 8 7.250000+6 8.640000-2 8.000000+6 8.640000-2 9.000000+6 8.640000-2803440 4 9 1.000000+7 8.640000-2 1.100000+7 8.640000-2 1.200000+7 8.640000-2803440 4 10 1.300000+7 8.640000-2 1.325000+7 3.640000-2 1.375000+7 6.400000-3803440 4 11 1.400000+7 6.400000-3 1.450000+7 3.640000-2 1.500000+7 3.640000-2803440 4 12 1.525000+7 8.640000-2 1.600000+7 8.640000-2 1.650000+7 2.464000-1803440 4 13 1.700000+7 2.464000-1 1.800000+7 2.464000-1 1.900000+7 2.464000-1803440 4 14 2.000000+7 0.000000+0 803440 4 15 0.000000+0 0.000000+0 0 1 6 3803440 4 16 1.000000-5 0.000000+0 5.250000+5 3.600000-3 2.000000+7 0.000000+0803440 4 17 803440 099999 8034 0 0 0 0 0 0 0 8.22040E+4 2.02220E+2 0 0 34 108225 1451 1 0.0 0.0 0 0 0 68225 1451 2 1.00000E+0 2.00000E+7 0 0 10 20028225 1451 3 3.00000E+2 0.0 1 0 147 58225 1451 4 82-Pb-204 FEI EVAL-Oct01 K.I.Zolotarev 8225 1451 5 DIST-Feb2004 8225 1451 6 ----IRDF-2002 MATERIAL 8225 8225 1451 7 -----INCIDENT NEUTRON DATA 8225 1451 8 ------ENDF-6 FORMAT 8225 1451 9 ***************************************************************** 8225 1451 10 82-Pb-204 FEI EVAL-Oct01 K.I.Zolotarev 8225 1451 11 DIST-Oct01 20011014 8225 1451 12 ----BROND-2 MATERIAL 8225 8225 1451 13 ------Russian Reactor Dosimetry File RRDF-2002 8225 1451 14 ***************************************************************** 8225 1451 15 ----- MF=3 MT=71 ----- 8225 1451 16 For the IRDF-2002 file this reaction was converted at IAEA/NDS 8225 1451 17 The reaction MF/MT=3/71 was converted to MF/MT=10/ 4 8225 1451 18 The corresponding co-variance files were also converted 8225 1451 19 The reaction MF/MT=33/71 was converted to MF/MT=40/ 4 8225 1451 20 ***************************************************************** 8225 1451 21 Author of evaluation: K.I.Zolotarev 8225 1451 22 ***************************************************************** 8225 1451 23 8225 1451 24 ----- MF=3 MT=71 ----- 8225 1451 25 8225 1451 26 In the section MT= 71 are given evaluated cross section data 8225 1451 27 of neutron inelastic excitation of the 67.2-min isomeric state in 8225 1451 28 in the lead-204 in the energy range from threshold to 20 MeV. 8225 1451 29 The isomeric state is the 21-excited level in Pb-204. This level 8225 1451 30 has energy 2185.79 keV with spin and parity 9- [1]. 8225 1451 31 Data base for the evaluation Pb-204(n,n')Pb-204m excitation 8225 1451 32 function was formed from microscopic experimental data [2-9] and 8225 1451 33 data received from theoretical model calculations. Experimental 8225 1451 34 data [3-9] were corrected to the new standards. Data of Decowski 8225 1451 35 et al. [6] were renormalized also to the result of precise measu- 8225 1451 36 rements of Ryves et al. at 14.3 MeV [8]. Uncertainty for cross 8225 1451 37 section value measured by Bornemisza-Pauspertl et al. at 2.80 MeV 8225 1451 38 [2] was evaluated as +-35 %. Theoretical model calculations was 8225 1451 39 done by means of GNASH code [10]. New modified version of GNASH 8225 1451 40 was used for calculations [11]. 8225 1451 41 The evaluation Pb-204(n,n')Pb-204m excitation function from 8225 1451 42 threshold to 20 MeV has been carried out within the framework of 8225 1451 43 generalized least squares method , rational function was used as 8225 1451 44 model function [12]. Procedure of calculation recommended cross 8225 1451 45 section data was performed by means of PADE-2 code [13]. 8225 1451 46 Integral experimental data for U-235 neutron fission spectrum 8225 1451 47 [14,15] and Cf-252 spontaneous fission neutron spectrum [16,17] 8225 1451 48 was used for testing evaluated Pb-204(n,n')Pb-204m excitation fun-8225 1451 49 ction. Data for U-235 thermal fission neutron spectrum and Cf-252 8225 1451 50 spontaneous fission neutron spectrum were taken from ref.[18] and 8225 1451 51 [19], respectively. The results of testing are given in the 8225 1451 52 table 1. 8225 1451 53 Table 1. 8225 1451 54 --------------------------------------------------------------- 8225 1451 55 TYPE OF SPECTRUM I ,MB (calc.) I , MB (measured) 8225 1451 56 ----------------------I-----------------I---------------------- 8225 1451 57 I I 8225 1451 58 U-235 neutron fission I 17.770 I 18.900 +- 2.000 [14] 8225 1451 59 I I 19.010 +- 1.536 [15] 8225 1451 60 I I 8225 1451 61 ----------------------I-----------------I---------------------- 8225 1451 62 I I 8225 1451 63 CF-252 spont. fission I 20.373 I 20.900 +- 1.202 [16] 8225 1451 64 I I 20.850 +- 0.920 [17] 8225 1451 65 --------------------------------------------------------------- 8225 1451 66 8225 1451 67 ----- MF=33 MT=71 ----- 8225 1451 68 Uncertainties in the evaluated excitation function for the 8225 1451 69 reaction Pb-204(n,n')Pb-204m are given in the form of relative 8225 1451 70 covariance matrix for the 32-neutron energy groups (LB=5). Cova- 8225 1451 71 riance matrix of uncertainties was calculated simultaneously with 8225 1451 72 recommended cross section data by means of PADE-2 code. 8225 1451 73 Eigenvalues of the 6-th digits relative covariance matrix 8225 1451 74 given in the 33-file are the following: 8225 1451 75 2.56972E-09 3.22598E-09 3.49414E-09 4.00985E-09 8225 1451 76 4.91484E-09 5.78671E-09 6.70005E-09 8.71427E-09 8225 1451 77 1.03160E-08 1.33154E-08 1.55381E-08 2.07583E-08 8225 1451 78 2.40281E-08 3.05913E-08 4.05801E-08 4.76532E-08 8225 1451 79 5.93691E-08 9.33863E-08 1.58183E-07 2.72581E-06 8225 1451 80 1.76176E-04 7.87730E-04 1.22783E-03 2.11803E-03 8225 1451 81 3.22408E-03 4.16999E-03 4.89344E-03 1.08231E-02 8225 1451 82 2.01870E-02 3.34051E-02 7.41942E-02 8225 1451 83 8225 1451 84 References : 8225 1451 85 1. R.B.Firestone Table of Isotopes, Eighth edition, Vol. 1, 8225 1451 86 John Wiley & Sons, Inc., New York, 1995 8225 1451 87 2. P.Bornemisza-Pauspertl, J.Karolyi, G.Peto Atomki Kozl., v.10, 8225 1451 88 no.2, p.112, July 1968 8225 1451 89 3. J.Csikai, G.Peto Acta Phys. Hung., v.23, p.87, May 1967 8225 1451 90 4. A.K.Hankla, R.W.Fink, J.H.Hamilton Nucl. Phys., v.A180, 8225 1451 91 p.157, January 1972 8225 1451 92 5. G.N.Maslov, F.Nasyrov, N.F.Pashkin Yadernye Konstanty, v.9, 8225 1451 93 p.50, Obninsk, 1972 8225 1451 94 6. P.Decowski et al. Nucl. Phys., v.A204, p.121, April 1973 8225 1451 95 7. D.L.Smith, J.W.Meadows Report ANL-NDM-37, December 1977 8225 1451 96 8. T.B.Ryves, P.Kolkowski, A.C.Hooley Annals Nucl. Energy, 8225 1451 97 v.17, p.107, 1990 8225 1451 98 9. I.Kimura, K.Kobayashi Nucl. Sci. Eng., v.106, p.332, 1990 8225 1451 99 10. P.G.Young, E.D.Arthur A Preequilibrium Statistical Nuclear 8225 1451 100 Model Code for Calculation of Cross Section and Emission 8225 1451 101 Spectra. Report LA-6947, Los Alamos, 1977 8225 1451 102 11. E.L.Trykov, G.Ya.Tertychnyi Private communication, IPPE, 8225 1451 103 Obninsk, May 1999 8225 1451 104 12. S.Badikov,N.Rabotnov,K.Zolotarev Proc. of NEANSC Specialist's 8225 1451 105 Meeting on Evaluation and Processing of Covariance Data, Oak 8225 1451 106 Ridge , USA, 7-9 September 1992, OECD, Paris, 1993, p.105 8225 1451 107 13. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 8225 1451 108 14. I.Kimura, K.Kobayashi, T.Shibata Nucl. Sci. Technol., v.8, 8225 1451 109 p.59, February 1971 8225 1451 110 15. A.K.Brodskaja et al. Jadernyje Konstanty, no.23, v.4, 8225 1451 111 October 1976 8225 1451 112 16. J.Csikai, Z.Dezso Proc. Educational Seminar on the use of 8225 1451 113 Cf-252, Karlsruhe, 14-18 April 1975, p.29 8225 1451 114 17. K.Kobayashi et al. Report KURRI-AR-17, p.15, 1984 ; 8225 1451 115 K.Kobayashi et al. Progress Report NEANDC(J)-106/U, p.41, 8225 1451 116 September 1984 8225 1451 117 18. L.W.Weston et al. Evaluated Neutron Data File for U-235, 8225 1451 118 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 8225 1451 119 19. W.Mannhart Report IAEA-TECDOC-410, IAEA, Vienna, 1987, p158 8225 1451 120 ***************************************************************** 8225 1451 121 File 2 added to the pointwise file containing only the effective 8225 1451 122 scattering radius with no resonance parameters given. 8225 1451 123 Taken from JENDL-3.2 8225 1451 124 8225 1451 125 ***************************************************************** 8225 1451 126 ***************** Program LINEAR (VERSION 2002-1) ***************8225 1451 127 For All Data Greater than 1.0000E-10 barns in Absolute Value 8225 1451 128 Data Linearized to Within an Accuracy of .100000000 per-cent 8225 1451 129 ***************** Program SIGMA1 (VERSION 2002-1) ***************8225 1451 130 Data Doppler Broadened to 300.000000 Kelvin 8225 1451 131 for All Data Greater than 1.0000E-10 barns in Absolute Value 8225 1451 132 Data Linearized to Within an Accuracy pf .100000000 per-cent 8225 1451 133 ***************** Program FIXUP (Version 2002-1) ****************8225 1451 134 Corrected ZA/AWR in All Sections-----------------------------Yes 8225 1451 135 Corrected Thresholds-----------------------------------------Yes 8225 1451 136 Extended Cross Sections to 20 MeV----------------------------No 8225 1451 137 Allow Cross Section Deletion---------------------------------No 8225 1451 138 Allow Cross Section Reconstruction---------------------------No 8225 1451 139 Make All Cross Sections Non-Negative-------------------------Yes 8225 1451 140 Delete Energies Not in Ascending Order-----------------------Yes 8225 1451 141 Deleted Duplicate Points-------------------------------------Yes 8225 1451 142 Check for Ascending MAT/MF/MT Order--------------------------Yes 8225 1451 143 Check for Legal MF/MT Numbers--------------------------------Yes 8225 1451 144 Allow Creation of Missing Sections---------------------------No 8225 1451 145 Allow Insertion of Energy Points-----------------------------No 8225 1451 146 Create Uniform Energy Grid-----------------------------------No 8225 1451 147 Delete Section if Cross Section =0 at All Energies-----------Yes 8225 1451 148 ***************** Program GROUPIE (VERSION 2002-1) **************8225 1451 149 Unshielded Group Averages Using 640 Groups 8225 1451 150 Weighting Spectrum: Flat (Constant) Spectrum 8225 1451 151 1 451 156 18225 1451 152 2 151 4 18225 1451 153 8 4 2 18225 1451 154 10 4 63 18225 1451 155 40 4 98 18225 1451 156 8225 1 099999 8225 0 0 0 8.22040E+4 2.02221E+2 0 0 1 08225 2151 1 8.22040+ 4 1.00000+ 0 0 0 1 08225 2151 2 1.00000- 5 5.00000+ 4 0 0 0 08225 2151 3 0.00000+ 0 8.50000- 1 0 0 0 08225 2151 4 8225 2 099999 8225 0 0 0 8.22040E+4 2.02220E+2 0 0 1 18225 8 4 1 8.22040E+4 0.000000+0 10 0 0 08225 8 4 2 8225 8 099999 8225 0 0 0 8.22040E+4 2.02220E+2 0 0 1 0822510 4 1 0.0 -2.18579E+6 82204 0 1 180822510 4 2 180 1 822510 4 3 2100000.00 1.34936E-8 2200000.00 3.51923E-5 2300000.00 .000189196822510 4 4 2400000.00 .001283871 2500000.00 .004581835 2600000.00 .008006575822510 4 5 2700000.00 .011515626 2800000.00 .015279183 2900000.00 .019236602822510 4 6 3000000.00 .023323413 3100000.00 .027458638 3200000.00 .031593863822510 4 7 3300000.00 .035729088 3400000.00 .039819900 3500000.00 .043866300822510 4 8 3600000.00 .047857600 3700000.00 .051793800 3800000.00 .055666950822510 4 9 3900000.00 .059477050 4000000.00 .063287150 4100000.00 .067032108822510 4 10 4200000.00 .070711925 4300000.00 .074391742 4400000.00 .078071558822510 4 11 4500000.00 .081751375 4600000.00 .085431192 4700000.00 .089118290822510 4 12 4800000.00 .092812670 4900000.00 .096507050 5000000.00 .100201430822510 4 13 5100000.00 .103895810 5200000.00 .107668750 5300000.00 .111520250822510 4 14 5400000.00 .115371750 5500000.00 .119223250 5600000.00 .123175167822510 4 15 5700000.00 .127227500 5800000.00 .131279833 5900000.00 .135442167822510 4 16 6000000.00 .139714500 6100000.00 .143986833 6200000.00 .148383667822510 4 17 6300000.00 .152905000 6400000.00 .157426333 6500000.00 .162078000822510 4 18 6600000.00 .166860000 6700000.00 .171642000 6800000.00 .176562375822510 4 19 6900000.00 .181621125 7000000.00 .186679875 7100000.00 .191738625822510 4 20 7200000.00 .196883429 7300000.00 .202114286 7400000.00 .207345143822510 4 21 7500000.00 .212576000 7600000.00 .217806857 7700000.00 .223037714822510 4 22 7800000.00 .228268571 7900000.00 .233326500 8000000.00 .238211500822510 4 23 8100000.00 .243096500 8200000.00 .247725500 8300000.00 .252098500822510 4 24 8400000.00 .256166250 8500000.00 .259928750 8600000.00 .263295750822510 4 25 8700000.00 .266267250 8800000.00 .268757750 8900000.00 .270767250822510 4 26 9000000.00 .272370000 9100000.00 .273276000 9200000.00 .273592000822510 4 27 9300000.00 .273302000 9400000.00 .272398000 9500000.00 .270881000822510 4 28 9600000.00 .268761000 9700000.00 .266057500 9800000.00 .262672250822510 4 29 9900000.00 .258890750 10000000.0 .254652250 10100000.0 .249956750822510 4 30 10200000.0 .244904750 10300000.0 .239496250 10400000.0 .233798500822510 4 31 10500000.0 .227811500 10600000.0 .221824500 10700000.0 .215712667822510 4 32 10800000.0 .209476000 10900000.0 .203239333 11000000.0 .197002667822510 4 33 11100000.0 .190766000 11200000.0 .184529333 11300000.0 .178482167822510 4 34 11400000.0 .172624500 11500000.0 .166766833 11600000.0 .161108750822510 4 35 11700000.0 .155650250 11800000.0 .150372000 11900000.0 .145274000822510 4 36 12000000.0 .140363250 12100000.0 .135639750 12200000.0 .131102750822510 4 37 12300000.0 .126752250 12400000.0 .122582250 12500000.0 .118592750822510 4 38 12600000.0 .114774500 12700000.0 .111127500 12800000.0 .107640500822510 4 39 12900000.0 .104313500 13000000.0 .101134025 13100000.0 .098102075822510 4 40 13200000.0 .095205225 13300000.0 .092443475 13400000.0 .089804450822510 4 41 13500000.0 .087288150 13600000.0 .084882850 13700000.0 .082588550822510 4 42 13800000.0 .080394125 13900000.0 .078299575 14000000.0 .076294725822510 4 43 14100000.0 .074379575 14200000.0 .072544700 14300000.0 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1.450000+7 1.500000+7 1.600000+7 1.700000+7822540 4 9 1.800000+7 1.900000+7 2.000000+7 0.000000+0 0.000000+0 0.000000+0822540 4 10 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0822540 4 11 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0822540 4 12 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0822540 4 13 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0822540 4 14 0.000000+0 0.000000+0 0.000000+0 0.000000+0 0.000000+0 6.053160-3822540 4 15 4.115730-3 3.116670-3 2.961290-3 3.015440-3 3.068430-3 3.122180-3822540 4 16 3.191450-3 3.270380-3 3.330530-3 3.323760-3 3.189580-3 2.873690-3822540 4 17 2.362520-3 1.719920-3 1.085450-3 6.067590-4 3.537590-4 2.953430-4822540 4 18 3.452160-4 4.185990-4 4.610410-4 4.510900-4 3.914730-4 2.986660-4822540 4 19 1.946180-4 7.102620-5 3.340260-5 2.090940-4 5.777680-4 9.758000-4822540 4 20 3.652730-3 3.188180-3 2.741690-3 2.500020-3 2.391720-3 2.370940-3822540 4 21 2.404200-3 2.459540-3 2.501280-3 2.488020-3 2.375450-3 2.128100-3822540 4 22 1.741980-3 1.267890-3 8.067540-4 4.607600-4 2.752690-4 2.260360-4822540 4 23 2.528310-4 2.982150-4 3.260610-4 3.223460-4 2.886880-4 2.352530-4822540 4 24 1.753850-4 1.054310-4 8.819490-5 1.980940-4 4.282670-4 7.023930-4822540 4 25 4.090330-3 3.809350-3 3.084810-3 2.409280-3 1.945010-3 1.713110-3822540 4 26 1.675720-3 1.764700-3 1.891470-3 1.955560-3 1.864770-3 1.574250-3822540 4 27 1.129950-3 6.693000-4 3.455270-4 2.288620-4 2.783260-4 3.939740-4822540 4 28 4.826620-4 4.913870-4 4.094080-4 2.562070-4 6.726690-5-1.174300-4822540 4 29 -2.960370-4-2.652780-4 1.334280-4 7.404230-4 9.977190-4 4.192280-3822540 4 30 3.615200-3 2.756360-3 2.026880-3 1.585040-3 1.448570-3 1.554920-3822540 4 31 1.788280-3 1.997760-3 2.028940-3 1.784090-3 1.295080-3 7.395430-4822540 4 32 3.407490-4 2.168620-4 3.235160-4 5.202110-4 6.678800-4 6.838410-4822540 4 33 5.497210-4 2.955960-4-2.127310-5-3.347490-4-6.455630-4-6.120310-4822540 4 34 4.206840-5 1.036910-3 1.369200-3 3.337220-3 2.715200-3 2.122560-3822540 4 35 1.728440-3 1.573850-3 1.622850-3 1.787280-3 1.942730-3 1.955000-3822540 4 36 1.731610-3 1.287780-3 7.683660-4 3.690520-4 2.055300-4 2.548860-4822540 4 37 4.047030-4 5.353800-4 5.700850-4 4.849860-4 2.980360-4 5.201480-5822540 4 38 -2.007100-4-4.671060-4-4.766410-4 1.753560-5 8.191410-4 1.160780-3822540 4 39 2.441450-3 2.138780-3 1.913630-3 1.799600-3 1.783420-3 1.818670-3822540 4 40 1.835170-3 1.754800-3 1.523250-3 1.151680-3 7.327830-4 3.941010-4822540 4 41 2.161850-4 1.927320-4 2.578620-4 3.363690-4 3.757390-4 3.541240-4822540 4 42 2.748900-4 1.572480-4 2.758470-5-1.228700-4-1.583610-4 7.210250-5822540 4 43 5.022620-4 7.992990-4 2.109570-3 2.066140-3 2.021340-3 1.971810-3822540 4 44 1.901400-3 1.785200-3 1.597680-3 1.328150-3 9.989460-4 6.686560-4822540 4 45 4.053520-4 2.473360-4 1.879950-4 1.923760-4 2.219010-4 2.487100-4822540 4 46 2.586200-4 2.484210-4 2.218820-4 1.863140-4 1.365060-4 1.095370-4822540 4 47 1.667880-4 3.241550-4 5.620980-4 2.162730-3 2.194410-3 2.155060-3822540 4 48 2.038730-3 1.841680-3 1.568570-3 1.241280-3 9.033450-4 6.092850-4822540 4 49 3.991440-4 2.794760-4 2.286070-4 2.165530-4 2.199510-4 2.258760-4822540 4 50 2.293520-4 2.297220-4 2.280340-4 2.256590-4 2.235460-4 2.278300-4822540 4 51 2.511290-4 3.214710-4 5.141590-4 2.294750-3 2.304690-3 2.209580-3822540 4 52 2.000860-3 1.685430-3 1.297840-3 9.040340-4 5.809970-4 3.773200-4822540 4 53 2.890450-4 2.741090-4 2.853540-4 2.918090-4 2.816490-4 2.560570-4822540 4 54 2.222490-4 1.886790-4 1.625550-4 1.497200-4 1.850350-4 2.819620-4822540 4 55 4.279470-4 6.049040-4 2.388890-3 2.371990-3 2.220610-3 1.918710-3822540 4 56 1.490500-3 1.015220-3 6.062410-4 3.509160-4 2.609900-4 2.810290-4822540 4 57 3.363160-4 3.716400-4 3.626150-4 3.095780-4 2.269990-4 1.346430-4822540 4 58 5.186300-5-1.528340-5 2.768240-5 2.366830-4 5.416980-4 7.380010-4822540 4 59 2.467190-3 2.428130-3 2.199940-3 1.775320-3 1.228460-3 7.049070-4822540 4 60 3.457620-4 2.030410-4 2.266640-4 3.197410-4 3.976080-4 4.138000-4822540 4 61 3.586830-4 2.474170-4 1.074840-4-3.066840-5-1.670910-4-1.540770-4822540 4 62 1.281860-4 5.757270-4 8.242120-4 2.528710-3 2.427200-3 2.072030-3822540 4 63 1.508680-3 8.913110-4 4.070610-4 1.622030-4 1.370690-4 2.335570-4822540 4 64 3.469890-4 4.076110-4 3.881820-4 2.944740-4 1.518610-4-6.200680-6822540 4 65 -1.906720-4-2.405740-4 2.059410-5 5.092830-4 8.361200-4 2.481690-3822540 4 66 2.270140-3 1.789640-3 1.165110-3 5.959140-4 2.350450-4 1.130070-4822540 4 67 1.604110-4 2.761000-4 3.777720-4 4.198440-4 3.906200-4 3.019850-4822540 4 68 1.787130-4-1.924600-7-1.140470-4 3.356710-5 4.377700-4 8.644110-4822540 4 69 2.257900-3 1.980930-3 1.497860-3 9.664880-4 5.494970-4 3.256820-4822540 4 70 2.785150-4 3.407000-4 4.403740-4 5.251840-4 5.676180-4 5.611320-4822540 4 71 5.135370-4 4.034660-4 2.809920-4 3.093840-4 5.829890-4 1.137240-3822540 4 72 2.000580-3 1.826280-3 1.520870-3 1.195250-3 9.411980-4 7.950340-4822540 4 73 7.457830-4 7.605310-4 8.046270-4 8.515680-4 8.855910-4 9.006540-4822540 4 74 8.918920-4 8.697480-4 9.247820-4 1.198590-3 1.930540-3 2.069600-3822540 4 75 2.170730-3 2.127140-3 1.982240-3 1.792810-3 1.603980-3 1.442320-3822540 4 76 1.319490-3 1.238160-3 1.196730-3 1.192270-3 1.252010-3 1.444090-3822540 4 77 1.791820-3 2.368020-3 3.354430-3 2.757910-3 3.136490-3 3.244060-3822540 4 78 3.107020-3 2.801640-3 2.414680-3 2.021720-3 1.679890-3 1.427830-3822540 4 79 1.288530-3 1.324970-3 1.778140-3 2.659800-3 3.857740-3 5.178410-3822540 4 80 3.951380-3 4.381340-3 4.382000-3 4.025400-3 3.439110-3 2.759100-3822540 4 81 2.104450-3 1.567490-3 1.212250-3 1.125220-3 1.782580-3 3.259190-3822540 4 82 5.214970-3 6.909040-3 5.098540-3 5.270890-3 4.939470-3 4.244030-3822540 4 83 3.358470-3 2.449770-3 1.657290-3 1.085100-3 8.215140-4 1.550350-3822540 4 84 3.460510-3 6.039040-3 8.070540-3 5.590640-3 5.343050-3 4.654550-3822540 4 85 3.706320-3 2.684930-3 1.754210-3 1.042510-3 6.194140-4 1.274620-3822540 4 86 3.314340-3 6.172660-3 8.424730-3 5.207660-3 4.630850-3 3.773000-3822540 4 87 2.808770-3 1.897690-3 1.169280-3 6.629810-4 1.141720-3 2.986270-3822540 4 88 5.708020-3 8.000890-3 4.237350-3 3.594190-3 2.838430-3 2.099130-3822540 4 89 1.484160-3 1.003900-3 1.272530-3 2.672740-3 4.889680-3 7.008540-3822540 4 90 3.241590-3 2.798940-3 2.346960-3 1.954760-3 1.616460-3 1.713150-3822540 4 91 2.545590-3 4.012270-3 5.735180-3 2.716800-3 2.616640-3 2.520100-3822540 4 92 2.426860-3 2.449610-3 2.727590-3 3.350830-3 4.472360-3 2.879730-3822540 4 93 3.109940-3 3.340090-3 3.428280-3 3.287320-3 3.123280-3 3.473200-3822540 4 94 3.658080-3 4.259390-3 4.573780-3 4.243340-3 3.475180-3 2.932480-3822540 4 95 5.424620-3 6.383490-3 6.350240-3 5.263230-3 3.318230-3 8.625060-3822540 4 96 1.000150-2 9.656750-3 6.041350-3 1.365320-2 1.538780-2 1.111050-2822540 4 97 1.989260-2 1.732600-2 2.174600-2 822540 4 98 822540 099999 8225 0 0 0 0 0 0 0 9.02320E+4 2.30044E+2 0 1 34 109040 1451 1 0.0 1.00000E+0 0 0 0 69040 1451 2 1.00000E+0 2.00000E+7 0 0 10 20029040 1451 3 3.00000E+2 0.0 1 0 208 69040 1451 4 90-Th-232 BNL,ANL+ EVAL-DEC77 BHAT,SMITH,LEONARD,DESAUSSURE+ 9040 1451 5 DIST-Feb2004 9040 1451 6 ----IRDF-2002 MATERIAL 9040 9040 1451 7 -----INCIDENT NEUTRON DATA 9040 1451 8 ------ENDF-6 FORMAT 9040 1451 9 ******************************************************************9040 1451 10 90-TH-232 BNL,ANL+ EVAL-DEC77 BHAT,SMITH,LEONARD,DESAUSSURE+ 9040 1451 11 DIST-FEB90 19900205 9040 1451 12 ----ENDF/B-VI MATERIAL 9040 9040 1451 13 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9040 1451 14 DOSIMETRY 9040 1451 15 ******************************************************************9040 1451 16 ENDF/B-V MATERIAL CONVERTED TO ENDF-6 FORMAT BY NNDC 9040 1451 17 * * * * * * *9040 1451 18 9040 1451 19 PRINCIPAL EVALUATORS- J.W.MEADOWS,W.P.POENITZ,A.B.SMITH,D.L.SMITH9040 1451 20 AND J.F.WHALEN(ANL) 9040 1451 21 R.J.HOWERTON(LLL) 9040 1451 22 B.R.LEONARD ET.AL.(BNW) 9040 1451 23 G.DE SAUSSURE,R.L.MACKLIN AND R.GWIN(ORNL) 9040 1451 24 D.K. OLSEN (ORNL) 9040 1451 25 THIS EVALUATION ASSEMBLED BY M.R.BHAT(BNL) USING THE FOLLOWING 9040 1451 26 EVALUATIONS 9040 1451 27 9040 1451 28 NU-BAR(PROMPT) - R.GWIN(ORNL)(REF 4) 9040 1451 29 DELAYED NEUTRONS -M.BRADY AND T.ENGLAND (REF 5) 9040 1451 30 ENERGY RELEASE IN FISSION - R.SHER ET.AL.(STANFORD)(REF 6) 9040 1451 31 THERMAL RANGE - B.R.LEONARD JR.ET.AL.(REF 2) WITH CHANGES BY9040 1451 32 M.R.BHAT(REF 31) 9040 1451 33 RESOLVED RESONANCE REGION - D.K. OLSEN (REF 37) 9040 1451 34 9040 1451 35 UNRESOLVED RESONANCE REGION - G.DE SAUSSURE AND R.L.MACKLIN 9040 1451 36 (REF 3)WITH CHANGES BY M.R.BHAT(REF 31) 9040 1451 37 SMOOTH CROSS-SECTIONS 50KEV TO 20MEV - 9040 1451 38 TOTAL - A.B.SMITH(ANL)(REF1) 9040 1451 39 FISSION - W.P.POENITZ(ANL)(REF1) 9040 1451 40 CAPTURE - W.P.POENITZ(ANL)(REF1) 9040 1451 41 INELASTIC AND OTHER CROSS-SECTIONS - J.W.MEADOWS ET.AL.(ANL)9040 1451 42 (REF1)9040 1451 43 FISSION NEUTRON SPECTRUM - M.R.BHAT(BNL) (REF 31) 9040 1451 44 GAMMA-RAY PRODUCTION - R.J.HOWERTON(LLL) 9040 1451 45 9040 1451 46 MF = 1 9040 1451 47 9040 1451 48 MT=452(NU-BAR TOTAL) CONSISTENT WITH MT=455 AND MT=456 9040 1451 49 MT=455(DELAYED NEUTRON YIELDS) (REF 5) 9040 1451 50 MT=456(PROMPT NU-BAR) BASED ON CF-252 NU-BAR(PROMPT)=3.757 9040 1451 51 EVALUATION BY GWIN(ORNL) BASED ON A STRAIGHT LINE LEAST- 9040 1451 52 SQUARES FIT TO THE AVAILABLE DATA LISTED IN REF.4. 9040 1451 53 MT=458(ENERGY RELEASE IN FISSION)EVALUATION BY R.SHER ET.AL(STAN-9040 1451 54 FORD)(REF 6) 9040 1451 55 9040 1451 56 MF = 2 9040 1451 57 9040 1451 58 RESOLVED RESONANCE REGION - D.K. OLSEN (REF 37) RESONANCE 9040 1451 59 PARAMETERS ARE UNCERTAINTY WEIGHTED AVERAGE OF AVAILABLE 9040 1451 60 DATA AND THE TWO BOUND RESONANCES OF REF 2 WITH MF=3 FILES 9040 1451 61 FOR TRUNCATION, MISSED P-WAVES, 1/V CAPTURE, AND SMOOTH 9040 1451 62 CONNECTION TO THE END OF THE THERMAL REGION CROSS SECTIONS 9040 1451 63 AT 5 EV. AVERAGE CAPTURE WIDTH = 2.44 X 10-2 EV, S-WAVE 9040 1451 64 STRENGTH FUNCTION = 0.826 X 10-4, P-WAVE STRENGTH FUNCTION 9040 1451 65 ASSUMED = 1.6 X 10-4, 5 TO 4000 EV INFINITLY DILUTE CAPTURE 9040 1451 66 RESONANCE INTEGRAL = 81.58 B. 9040 1451 67 9040 1451 68 UNRESOLVED RESONANCE REGION - G.DE SAUSSURE AND R.L.MACKLIN(REF3)9040 1451 69 EVALUATION OF CAPTURE CROSS-SECTION USED FROM 4 TO 25KEV. 9040 1451 70 ABOVE 25KEV THIS WAS JOINED SMOOTHLY TO THE POENITZ EVALUAT-9040 1451 71 ION FROM 40KEV TO 50KEV. THESE WERE THEN FITTED USING THE 9040 1451 72 CODE UR(REF30) TO GIVE THE UNRESOLVED RESONANCE 9040 1451 73 PARAMETERS. IN THIS FIT THE S- AND D-WAVE PARAMETERS WERE 9040 1451 74 KEPT CONSTANT AT THE VALUES GIVEN BY DE SAUSSURE AND MACKLIN9040 1451 75 THE P-WAVE GAMMA WIDTH WAS INCREASED TO 2.52E-02EV AND THE 9040 1451 76 P-WAVE REDUCED NEUTRON WIDTH VARIED TO FIT THE CAPTURE 9040 1451 77 CROSS-SECTION. 9040 1451 78 THE CAPTURE CROSS-SECTION INPUT USED TO EXTRACT THE 9040 1451 79 UNRESOLVED RESONANCE PARAMETERS IN FILE 2/151 ARE GIVEN HERE9040 1451 80 FROM 4KEV TO 50KEV AT 1KEV INTERVAL. THESE ARE - 9040 1451 81 1.130,0.982,0.908,0.850,0.803,0.764,0.730,0.702,0.677,0.655,9040 1451 82 0.636,0.618,0.601,0.585,0.571,0.557,0.545,0.533,0.521,0.511,9040 1451 83 0.500,0.491,0.482,0.473,0.465,0.459,0.452,0.448,0.443,0.439,9040 1451 84 0.435,0.431,0.429,0.426,0.424,0.422,0.420,0.420,0.420,0.420,9040 1451 85 0.417,0.414,0.411,0.407,0.401,0.395,0.390. 9040 1451 86 9040 1451 87 MF = 3 9040 1451 88 9040 1451 89 THERMAL REGION(1.0E-05 TO 5.0EV) - MAINLY BASED ON LEONARD(REF 2)9040 1451 90 EVALUATION WITH THE FOLLOWING CHANGES. THE CAPTURE CROSS- 9040 1451 91 SECTION IN REF 2 NORMALISED TO 7.40B AT 2.53E-02 EV AND USED9040 1451 92 FROM 1.0E-05EV TO 2.53E-02EV. FROM 2.6E-02EV TO 14EV THE FIT9040 1451 93 TO THEIR CAPTURE DATA BY CHRIEN AND LIOU(REF 32) USED. THE 9040 1451 94 SCATTERING CROSS-SECTION WAS CHANGED TO ACCOMMODATE CHANGES 9040 1451 95 IN CAPTURE USING TOTAL SLIGHTLY MODIFIED FROM THAT GIVEN IN 9040 1451 96 REF 2. THESE SMALL CHANGES ARE DESCRIBED IN REF 31. 9040 1451 97 2.53E-02EV CROSS-SECTION CAPTURE 7.4 B 9040 1451 98 DILUTE CAPTURE RESONANCE INTEGRAL 86.1 B 9040 1451 99 (0.5 TO 2.0E+07 EV) 9040 1451 100 TOTAL CROSS-SECTION FROM 50KEV TO 20MEV - A.B.SMITH(ANL)(REF1) 9040 1451 101 MAINLY BASED ON REF9-13 FROM 0.025 TO 1 MEV,FROM 1 TO 5MEV 9040 1451 102 BASED ON REF 9-11,14-16.FROM 5 TO 15MEV EVALUATION BASED ON 9040 1451 103 REF11,15,17.FROM 15 TO 20MEV EVALUATION BASED ON NUCLEAR 9040 1451 104 MODEL CALCULATIONS DESCRIBED IN REF18 AND 19.TOTAL CROSS- 9040 1451 105 SECTION MODIFIED BY ABOUT 3PER-CENT BETWEEN 50 TO 150KEV 9040 1451 106 JOIN SMOOTHLY WITH THAT CALCULATED FROM THE UNRESOLVED 9040 1451 107 RESONANCE PARAMETERS BELOW 50KEV. 9040 1451 108 ELASTIC SCATTERING CROSS-SECTION FROM 50KEV TO 20MEV OBTAINED AS 9040 1451 109 THE DIFFERENCE BETWEEN TOTAL AND OTHER PARTIAL CROSSSECTIONS9040 1451 110 FISSION CROSS-SECTION UP TO 20MEV - W.P.POENITZ(REF1) DETAILS OF 9040 1451 111 THE EVALUATION AND THE DATA SETS USED ARE IN REF1. 9040 1451 112 FISSION CROSS-SECTION CHANGED TO BE CONSISTENT WITH 9040 1451 113 U-235(N,F) EVALUATION GIVEN IN REF 33 9040 1451 114 CAPTURE CROSS-SECTION 50KEV -20MEV - W.P.POENITZ. DETAILS OF THE 9040 1451 115 EVALUATION AND THE DATA SETS USED ARE GIVEN IN REF1. 9040 1451 116 CAPTURE CROSS-SECTION CHANGED TO BE CONSISTENT WITH 9040 1451 117 U-235(N,F) EVALUATION GIVEN IN REF 33 9040 1451 118 INELASTIC SCATTERING CROSS-SECTIONS - J.W.MEADOWS ET.AL.(REF 1) 9040 1451 119 EVALUATION BASED MAINLY ON THE DATA IN REF 20-25. INTER- 9040 1451 120 POLATED BETWEEN EXPERIMENTAL DATA AND EXTRAPOLATED TO 20MEV 9040 1451 121 USING COUPLED CHANNEL NUCLEAR MODEL CALCULATIONS OF REF26. 9040 1451 122 (N,2N) AND (N,3N) CROSS-SECTIONS - J.W.MEADOWS ET.AL.(REF1) 9040 1451 123 EXPERIMENTAL DATA USED(REF1) RENORMALISED TO A COMMON SET OF9040 1451 124 STANDARDS BEFORE EVALUATION. 9040 1451 125 9040 1451 126 MF = 33 9040 1451 127 MT = 18,102 BASED ON ERROR ESTIMATES IN THESE CROSS SECTIONS IN 9040 1451 128 REF. 32,1 9040 1451 129 9040 1451 130 PRINCIPAL REFERENCES 9040 1451 131 9040 1451 132 1 J.W.MEADOWS,W.P.POENITZ,A.B.SMITH,D.L.SMITH,J.F.WHALEN AND 9040 1451 133 R.J.HOWERTON ANL/NDM-35(FEB.1978) 9040 1451 134 2 D.F.NEWMAN,B.R.LEONARD JR.ET.AL. EPRI NP-222,MAY1977 9040 1451 135 3 G.DE SAUSSURE AND R.L.MACKLIN ORNL/TM-6161(ENDF-255)DEC.1977 9040 1451 136 4 R.GWIN ORNL/TM-6245(ENDF-262)MAY1978. 9040 1451 137 5 M.C.BRADY AND T.R.ENGLAND NSE 103,129(1989) 9040 1451 138 **DATA INSERTED INTO FILE AT BNL** 9040 1451 139 6. R.SHER + C.BECK EPRI 1771/81 + REV. 1/83 + PC TO MAGURNO 2/83 9040 1451 140 7 S.F.MUGHABGHAB AND D.I.GARBER,BNL-325(3RD EDN)1973 9040 1451 141 8 H.DERRIEN,NEANDC(E)163U(1975) 9040 1451 142 9 J.F.WHALEN AND A.B.SMITH,NUC.SCI.AND ENG.67,129(1978) 9040 1451 143 10J.MEADOWS,A.SMITH AND J.WHALEN,PRIV.COMM.1977 9040 1451 144 11U.FASOLI ET.AL.NUCL.PHYS.A151,369(1970) 9040 1451 145 12M.DIVADEENAM ET.AL. DISS.ABS.28,3834(1968) 9040 1451 146 13C.UTTLEY ET.AL.66PARIS,1,165,1966 9040 1451 147 14L.GREEN ET.AL.PROC.CONF.ON NUCL.CROSS-SECTIONS AND TECHNOLOGY, 9040 1451 148 KNOXVILLE,1,325(1971) 9040 1451 149 15D.FOSTER ET.AL.PRIV.COMM(1967) 9040 1451 150 16R.BATCHELOR ET.AL.NUCL.PHYS.65,236(1956) 9040 1451 151 17J.COON ET.AL.PHYS.REV.88,562(1952) 9040 1451 152 18P.GUENTHER ET.AL.ACCEPTED FOR PUBLICATION NUC.SCI.AND ENG.(19779040 1451 153 19C.PHILIS ET.AL. ANL/NDM-28(1977) 9040 1451 154 20A.SMITH PHYS.REV.126,718(1962) 9040 1451 155 21W.MCMURRAY ET.AL.SOUTHERN UNIV.NUCLEAR INST.REPORT,SUNI-41(19759040 1451 156 22J.HAOUAT ETAL.PROC.INER.CONF.ON INTERACTIONS OF NEUTRONS WITH 9040 1451 157 NUCLEI,CONF-760715(1976) 9040 1451 158 23R.BATCHELOR AND J.TOWLE,NUCL.PHYS.65,236(1965) 9040 1451 159 24R.BATCHELOR AND J.TOWLE,PROC.PHYS.SOC.73,193(1959) 9040 1451 160 25A.SMITH,PRIV.COMM,1970 AND 1977 9040 1451 161 26A.SMITH ET.AL.ACCEPTED FOR PUBLICATION NUC.SCI.AND ENG.(1977) 9040 1451 162 27M.SEGEV.ET.AL.TRANS.AM.NUCL.SOC,22,679(1975)AND PRIV.COMM(1976)9040 1451 163 28R.J.HOWERTON AND R.J.DOYAS,NUCL.SCI AND ENG,46,414(1971) 9040 1451 164 29YU.A.VASIL'EV ET.AL.PHYSICS OF NUCLEAR FISSION,N.A.PERFILOV AND9040 1451 165 V.P.EISMONT(EDS),ISRAEL PROGRAM OF SCI.TRANS.JERUSALEM(1964) 9040 1451 166 30E.M.PENNINGTON,PRIVATE COMMUNICATION(1973) 9040 1451 167 31M.R.BHAT,ENDF-268(TO BE PUBLISHED) 9040 1451 168 32R.E.CHRIEN AND H.LIOU(1978) TO BE PUBLISHED 9040 1451 169 33W.P.POENITZ,ANL/NDM-45(1978) TO BE PUBLISHED 9040 1451 170 34R.W.PEELE AND F.C.MAIENSCHEIN, NUCL.SCI.ENG. 40, 485 (1970) 9040 1451 171 35R.J. HOWERTON, D.E. CULLEN, R.C. HAIGHT, M.H. MACGREGOR, S.T. 9040 1451 172 PERKINS, AND E.F. PLECHARTY, "THE LLL EVALUATED NUCLEAR DATA 9040 1451 173 LIBRARY (ENDL): EVALUATION TECHNIQUES, REACTION INDEX, AND 9040 1451 174 DESCRIPTIONS OF INDIVIDUAL EVALUATIONS," UCRL-50400, VOL. 15, 9040 1451 175 PART A, LAWERENCE LIVERMORE LABORATORY (1975). 9040 1451 176 36S.T.PERKINS,R.C.HAIGHT AND R.J.HOWERTON,NUCL.SCI.ENG.57,1(1975)9040 1451 177 37D.K.OLSEN, ORNL/TM-8056(1982), ENDF-319. 9040 1451 178 ******************************************************************9040 1451 179 9040 1451 180 9040 1451 181 9040 1451 182 9040 1451 183 ******************************************************************9040 1451 184 ***************** Program LINEAR (VERSION 2002-1) ***************9040 1451 185 For All Data Greater than 1.0000E-10 barns in Absolute Value 9040 1451 186 Data Linearized to Within an Accuracy of .100000000 per-cent 9040 1451 187 ***************** Program RECENT (VERSION 2002-1) ***************9040 1451 188 for All Data Greater than 1.0000E-10 barns in Absolute Value 9040 1451 189 Data Linearized to within an Accuracy of .100000000 per-cent 9040 1451 190 ***************** Program SIGMA1 (VERSION 2002-1) ***************9040 1451 191 Data Doppler Broadened to 300.000000 Kelvin 9040 1451 192 for All Data Greater than 1.0000E-10 barns in Absolute Value 9040 1451 193 Data Linearized to Within an Accuracy pf .100000000 per-cent 9040 1451 194 ***************** Program FIXUP (Version 2002-1) ****************9040 1451 195 Corrected ZA/AWR in All Sections-----------------------------Yes 9040 1451 196 Corrected Thresholds-----------------------------------------Yes 9040 1451 197 Extended Cross Sections to 20 MeV----------------------------No 9040 1451 198 Allow Cross Section Deletion---------------------------------No 9040 1451 199 Allow Cross Section Reconstruction---------------------------No 9040 1451 200 Make All Cross Sections Non-Negative-------------------------Yes 9040 1451 201 Delete Energies Not in Ascending Order-----------------------Yes 9040 1451 202 Deleted Duplicate Points-------------------------------------Yes 9040 1451 203 Check for Ascending MAT/MF/MT Order--------------------------Yes 9040 1451 204 Check for Legal MF/MT Numbers--------------------------------Yes 9040 1451 205 Allow Creation of Missing Sections---------------------------No 9040 1451 206 Allow Insertion of Energy Points-----------------------------No 9040 1451 207 Create Uniform Energy Grid-----------------------------------No 9040 1451 208 Delete Section if Cross Section =0 at All Energies-----------Yes 9040 1451 209 ***************** Program GROUPIE (VERSION 2002-1) **************9040 1451 210 Unshielded Group Averages Using 640 Groups 9040 1451 211 Weighting Spectrum: Flat (Constant) Spectrum 9040 1451 212 1 451 218 19040 1451 213 2 151 4 09040 1451 214 3 18 72 09040 1451 215 3 102 217 09040 1451 216 33 18 9 09040 1451 217 33 102 7 09040 1451 218 9040 1 099999 9040 0 0 0 9.02320E+4 2.30045E+2 0 0 1 09040 2151 1 9.02320E+4 1.00000E+0 0 0 1 09040 2151 2 5.00000E+0 4.00000E+3 0 0 0 09040 2151 3 0.0 9.72000E-1 0 0 0 09040 2151 4 9040 2 099999 9040 0 0 0 9.02320E+4 2.30044E+2 0 0 0 09040 3 18 1 1.88470E+8 1.88470E+8 0 0 1 2069040 3 18 2 206 1 9040 3 18 3 500000.000 1.37500E-6 525000.000 4.12499E-6 550000.000 6.87498E-69040 3 18 4 575000.000 9.62500E-6 600000.000 1.44501E-5 630000.000 2.13500E-59040 3 18 5 660000.000 2.82500E-5 690000.000 4.12173E-5 720000.000 8.03773E-59040 3 18 6 760000.000 .000130050 800000.000 .000154600 840000.000 .0002595459040 3 18 7 880000.000 .000567766 920000.000 .001115025 960000.000 .0011350869040 3 18 8 1000000.00 .001805998 1100000.00 .003257634 1200000.00 .0100931179040 3 18 9 1300000.00 .038089744 1400000.00 .064670533 1500000.00 .0994199279040 3 18 10 1600000.00 .097810452 1700000.00 .084170147 1800000.00 .0952105919040 3 18 11 1900000.00 .116499192 2000000.00 .118500645 2100000.00 .1289995989040 3 18 12 2200000.00 .130499193 2300000.00 .117750737 2400000.00 .1157501309040 3 18 13 2500000.00 .118800036 2600000.00 .122399807 2700000.00 .1254663819040 3 18 14 2800000.00 .125333781 2900000.00 .129000649 3000000.00 .1374367209040 3 18 15 3100000.00 .138499460 3200000.00 .138262606 3300000.00 .1396000289040 3 18 16 3400000.00 .141199830 3500000.00 .141599831 3600000.00 .1408000329040 3 18 17 3700000.00 .140250187 3800000.00 .141200072 3900000.00 .1423999139040 3 18 18 4000000.00 .142899857 4100000.00 .142699906 4200000.00 .1424999559040 3 18 19 4300000.00 .142300003 4400000.00 .142100052 4500000.00 .1422143549040 3 18 20 4600000.00 .142642911 4700000.00 .143071468 4800000.00 .1435000259040 3 18 21 4900000.00 .143928582 5000000.00 .144517803 5100000.00 .1454997569040 3 18 22 5200000.00 .145249710 5300000.00 .143749811 5400000.00 .1417500209040 3 18 23 5500000.00 .139250337 5600000.00 .138500176 5700000.00 .1390002899040 3 18 24 5800000.00 .141500578 5900000.00 .148001097 6000000.00 .1620014499040 3 18 25 6100000.00 .182000830 6200000.00 .203750523 6300000.00 .2272505289040 3 18 26 6400000.00 .252499823 6500000.00 .276599003 6600000.00 .2977987789040 3 18 27 6700000.00 .318998553 6800000.00 .340198328 6900000.00 .3613981039040 3 18 28 7000000.00 .376248272 7100000.00 .384748835 7200000.00 .3904989219040 3 18 29 7300000.00 .393498528 7400000.00 .391373544 7500000.00 .3841239679040 3 18 30 7600000.00 .376874391 7700000.00 .369631916 7800000.00 .3644204229040 3 18 31 7900000.00 .360806729 8000000.00 .356899918 8100000.00 .3526999919040 3 18 32 8200000.00 .348500063 8300000.00 .344300135 8400000.00 .3401002089040 3 18 33 8500000.00 .336600256 8600000.00 .333800279 8700000.00 .3310003029040 3 18 34 8800000.00 .328200325 8900000.00 .325400348 9000000.00 .3236003459040 3 18 35 9100000.00 .322800318 9200000.00 .322000290 9300000.00 .3212002629040 3 18 36 9400000.00 .320400235 9500000.00 .320200157 9600000.00 .3206000309040 3 18 37 9700000.00 .320999903 9800000.00 .321399775 9900000.00 .3217996489040 3 18 38 10000000.0 .321099623 10100000.0 .319299700 10200000.0 .3174997779040 3 18 39 10300000.0 .315699855 10400000.0 .313899932 10500000.0 .3121000099040 3 18 40 10600000.0 .310300086 10700000.0 .308500164 10800000.0 .3067002419040 3 18 41 10900000.0 .304900318 11000000.0 .304000321 11100000.0 .3040002509040 3 18 42 11200000.0 .304000179 11300000.0 .304000107 11400000.0 .3040000369040 3 18 43 11500000.0 .304000033 11600000.0 .304000099 11700000.0 .3040001669040 3 18 44 11800000.0 .304000232 11900000.0 .304000298 12000000.0 .3048003069040 3 18 45 12100000.0 .306400257 12200000.0 .308000208 12300000.0 .3096001599040 3 18 46 12400000.0 .311200110 12500000.0 .313025081 12600000.0 .3150750729040 3 18 47 12700000.0 .317125064 12800000.0 .319175055 12900000.0 .3212250479040 3 18 48 13000000.0 .323275038 13100000.0 .325325030 13200000.0 .3273750219040 3 18 49 13300000.0 .329425013 13400000.0 .331475004 13500000.0 .3335501219040 3 18 50 13600000.0 .335650362 13700000.0 .337750604 13800000.0 .3398508469040 3 18 51 13900000.0 .341951087 14000000.0 .346751036 14100000.0 .3542506929040 3 18 52 14200000.0 .361750348 14300000.0 .369250003 14400000.0 .3767496599040 3 18 53 14500000.0 .384249315 14600000.0 .389874281 14700000.0 .3936245569040 3 18 54 14800000.0 .397374831 14900000.0 .401125106 15000000.0 .4053668939040 3 18 55 15100000.0 .410100194 15200000.0 .414833495 15300000.0 .4195667969040 3 18 56 15400000.0 .424300097 15500000.0 .429033398 15600000.0 .4337666999040 3 18 57 15700000.0 .438483313 15800000.0 .443099809 15900000.0 .4476996189040 3 18 58 16000000.0 .451299505 16100000.0 .453899472 16200000.0 .4564994389040 3 18 59 16300000.0 .459099404 16400000.0 .461699371 16500000.0 .4642993379040 3 18 60 16600000.0 .466899304 16700000.0 .469499270 16800000.0 .4720992369040 3 18 61 16900000.0 .474699203 17000000.0 .475649295 17100000.0 .4749495149040 3 18 62 17200000.0 .474249732 17300000.0 .473549950 17400000.0 .4728501699040 3 18 63 17500000.0 .472150387 17600000.0 .471450606 17700000.0 .4707508249040 3 18 64 17800000.0 .470051042 17900000.0 .469351261 18000000.0 .4713512799040 3 18 65 18100000.0 .476051098 18200000.0 .480750917 18300000.0 .4854507359040 3 18 66 18400000.0 .490150554 18500000.0 .494850373 18600000.0 .4995501929040 3 18 67 18700000.0 .504250010 18800000.0 .508949829 18900000.0 .5136496489040 3 18 68 19000000.0 .517499576 19100000.0 .520499613 19200000.0 .5234996509040 3 18 69 19300000.0 .526499687 19400000.0 .529499724 19500000.0 .5324997619040 3 18 70 19600000.0 .535499798 19700000.0 .538499835 19800000.0 .5414998729040 3 18 71 19900000.0 .544499909 20000000.0 0.0 9040 3 18 72 9040 3 099999 9.02320E+4 2.30044E+2 0 0 0 09040 3102 1 4.78640E+6 4.78640E+6 0 0 1 6419040 3102 2 641 1 9040 3102 3 .000100000 117.717596 .000105000 114.972465 .000110000 112.3645769040 3102 4 .000115000 109.961479 .000120000 107.152220 .000127500 104.0353799040 3102 5 .000135000 101.187548 .000142500 98.5704503 .000150000 95.73802059040 3102 6 .000160000 92.7873368 .000170000 90.1026352 .000180000 87.63829839040 3102 7 .000190000 85.3454282 .000200000 83.2519344 .000210000 81.28235679040 3102 8 .000220000 79.4530263 .000230000 77.7534639 .000240000 75.75853689040 3102 9 .000255000 73.5659823 .000270000 71.8742409 .000280000 69.99118079040 3102 10 .000300000 67.6943744 .000320000 65.6147172 .000340000 63.71014439040 3102 11 .000360000 61.9559112 .000380000 60.3543582 .000400000 58.68475839040 3102 12 .000425000 56.9764152 .000450000 55.4155706 .000475000 53.98450369040 3102 13 .000500000 52.6386001 .000525000 51.4083118 .000550000 50.24608029040 3102 14 .000575000 49.1651529 .000600000 48.0573418 .000630000 46.91773229040 3102 15 .000660000 45.8748715 .000690000 44.8744625 .000720000 43.81004259040 3102 16 .000760000 42.6686716 .000800000 41.6088204 .000840000 40.63725899040 3102 17 .000880000 39.7138277 .000920000 38.8674685 .000960000 38.05690769040 3102 18 .001000000 37.2166694 .001050000 36.3394262 .001100000 35.51694219040 3102 19 .001150000 34.7590912 .001200000 33.8650807 .001275000 32.88238449040 3102 20 .001350000 31.9838228 .001425000 31.1452380 .001500000 30.25449569040 3102 21 .001600000 29.3230418 .001700000 28.4726712 .001800000 27.68617629040 3102 22 .001900000 26.9677133 .002000000 26.3015200 .002100000 25.67635889040 3102 23 .002200000 25.1039807 .002300000 24.5566911 .002400000 23.93225549040 3102 24 .002550000 23.2382004 .002700000 22.6952333 .002800000 22.10554159040 3102 25 .003000000 21.3795011 .003200000 20.7185473 .003400000 20.11387849040 3102 26 .003600000 19.5622492 .003800000 19.0528074 .004000000 18.52179629040 3102 27 .004250000 17.9828720 .004500000 17.4910479 .004750000 17.03148329040 3102 28 .005000000 16.6099462 .005250000 16.2191227 .005500000 15.84959409040 3102 29 .005750000 15.5105884 .006000000 15.1545074 .006300000 14.79878309040 3102 30 .006600000 14.4627117 .006900000 14.1518952 .007200000 13.81281779040 3102 31 .007600000 13.4492601 .008000000 13.1166977 .008400000 12.80447929040 3102 32 .008800000 12.5138121 .009200000 12.2436716 .009600000 11.98660089040 3102 33 .010000000 11.7204259 .010500000 11.4390181 .011000000 11.18162339040 3102 34 .011500000 10.9360800 .012000000 10.6546914 .012750000 10.34322669040 3102 35 .013500000 10.0546958 .014250000 9.78988762 .015000000 9.507197929040 3102 36 .016000001 9.21021411 .017000001 8.93772944 .017999999 8.688957049040 3102 37 .018999999 8.46027176 .020000000 8.24568794 .021000000 8.049197419040 3102 38 .022000000 7.86409138 .023000000 7.69167626 .024000000 7.489465889040 3102 39 .025500000 7.27213432 .027000001 7.10017272 .028000001 6.913551159040 3102 40 .029999999 6.68370492 .032000002 6.47421280 .034000002 6.282990349040 3102 41 .035999998 6.10855617 .037999999 5.94753456 .039999999 5.779328459040 3102 42 .042500000 5.60831761 .045000002 5.45204742 .047499999 5.306473969040 3102 43 .050000001 5.17257368 .052499998 5.04852697 .055000000 4.930998419040 3102 44 .057500001 4.82340078 .059999999 4.71022229 .063000001 4.596858059040 3102 45 .066000000 4.48999040 .068999998 4.39041832 .071999997 4.282415399040 3102 46 .075999998 4.16667382 .079999998 4.06076698 .083999999 3.961125819040 3102 47 .088000000 3.86836194 .092000000 3.78228946 .096000001 3.699986769040 3102 48 .100000001 3.61485380 .104999997 3.52466255 .109999999 3.442182359040 3102 49 .115000002 3.36336348 .119999997 3.27302265 .127499998 3.173399779040 3102 50 .135000005 3.08098779 .142499998 2.99586987 .150000006 2.904812439040 3102 51 .159999996 2.80887282 .170000002 2.72093103 .180000007 2.640377819040 3102 52 .189999998 2.56634734 .200000003 2.49683305 .209999993 2.433374679040 3102 53 .219999999 2.37339008 .230000004 2.31735697 .239999995 2.252175989040 3102 54 .254999995 2.17907301 .270000011 2.12333892 .280000001 2.061599899040 3102 55 .300000012 1.98578742 .319999993 1.91625736 .340000004 1.852888159040 3102 56 .360000014 1.79473111 .379999995 1.74025445 .400000006 1.684892419040 3102 57 .425000012 1.62861092 .449999988 1.57516393 .474999994 1.526449959040 3102 58 .500000000 1.48180440 .524999976 1.43952136 .550000012 1.400689819040 3102 59 .574999988 1.36377395 .600000024 1.32608723 .629999995 1.288288039040 3102 60 .660000026 1.25244201 .689999998 1.21802661 .720000029 1.181535689040 3102 61 .759999990 1.14277962 .800000012 1.10644514 .839999974 1.073211149040 3102 62 .879999995 1.04172353 .920000017 1.01264674 .959999979 .9851716089040 3102 63 1.00000000 .956038107 1.04999995 .926183156 1.10000002 .8979742459040 3102 64 1.14999998 .872395836 1.20000005 .842109589 1.27499998 .8086236599040 3102 65 1.35000002 .778257586 1.42499995 .749757624 1.50000000 .7199226839040 3102 66 1.60000002 .688643210 1.70000005 .660468927 1.79999995 .6345953909040 3102 67 1.89999998 .611022622 2.00000000 .588899851 2.09999990 .5687421669040 3102 68 2.20000005 .550394153 2.29999995 .532947154 2.40000010 .5131081369040 3102 69 2.54999995 .491271807 2.70000005 .474584932 2.79999995 .4565969309040 3102 70 3.00000000 .434425625 3.20000005 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13.7818597 23.0000000 580.272945 24.0000000 2.930193819040 3102 84 25.5000000 .555913332 27.0000000 .309736817 28.0000000 .2130329379040 3102 85 30.0000000 .157303513 32.0000000 .132189787 34.0000000 .1188181339040 3102 86 36.0000000 .164088285 38.0000000 .135759158 40.0000000 .1265483069040 3102 87 42.5000000 .102371570 45.0000000 .157196272 47.5000000 .1233692639040 3102 88 50.0000000 .122941177 52.5000000 .179861901 55.0000000 .2502567379040 3102 89 57.5000000 90.2939115 60.0000000 .912830407 63.0000000 .6321927599040 3102 90 66.0000000 24.3366888 69.0000000 271.925256 72.0000000 .5815312739040 3102 91 76.0000000 .210880950 80.0000000 .142075695 84.0000000 .1178710979040 3102 92 88.0000000 .177175958 92.0000000 .103142449 96.0000000 .1451483209040 3102 93 100.000000 .156948948 105.000000 .177049679 110.000000 61.35374129040 3102 94 115.000000 .833155910 120.000000 52.4531072 127.500000 12.61246379040 3102 95 135.000000 .096591513 142.500000 .464631869 150.000000 .6322294319040 3102 96 160.000000 2.92064219 170.000000 40.5825351 180.000000 .1441355559040 3102 97 190.000000 34.1464386 200.000000 .313053176 210.000000 .3491740809040 3102 98 220.000000 25.1147614 230.000000 .154139326 240.000000 15.73711019040 3102 99 255.000000 11.5444632 270.000000 .185698563 280.000000 9.542452139040 3102 100 300.000000 8.64438446 320.000000 12.5427253 340.000000 8.447136309040 3102 101 360.000000 14.2073100 380.000000 .242211404 400.000000 3.519085549040 3102 102 425.000000 .087542524 450.000000 6.86556759 475.000000 5.928064139040 3102 103 500.000000 1.28130689 525.000000 3.35649766 550.000000 4.146925839040 3102 104 575.000000 2.75726884 600.000000 1.01066055 630.000000 3.533309959040 3102 105 660.000000 10.4884715 690.000000 4.60324876 720.000000 3.198388619040 3102 106 760.000000 1.35002046 800.000000 3.20385549 840.000000 3.029109269040 3102 107 880.000000 2.14459626 920.000000 1.92581674 960.000000 4.248621159040 3102 108 1000.00000 2.48660615 1050.00000 1.22423539 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.3526886039040 3102 134 57500.0000 .337769341 60000.0000 .324008124 63000.0000 .3119999999040 3102 135 66000.0000 .299999999 69000.0000 .288736879 72000.0000 .2780678019040 3102 136 76000.0000 .266022623 80000.0000 .256066783 84000.0000 .2480740989040 3102 137 88000.0000 .240299624 92000.0000 .234074339 96000.0000 .2281518469040 3102 138 100000.000 .222706480 105000.000 .218100000 110000.000 .2135000009040 3102 139 115000.000 .208900000 120000.000 .203365159 127500.000 .1984077539040 3102 140 135000.000 .193825772 142500.000 .189258045 150000.000 .1852126519040 3102 141 160000.000 .181541436 170000.000 .177744490 180000.000 .1728635829040 3102 142 190000.000 .167733061 200000.000 .163506608 210000.000 .1605000009040 3102 143 220000.000 .157500000 230000.000 .154500000 240000.000 .1510044389040 3102 144 255000.000 .148884890 270000.000 .147521763 280000.000 .1460064759040 3102 145 300000.000 .144700087 320000.000 .144100124 340000.000 .1435002079040 3102 146 360000.000 .142900291 380000.000 .142300374 400000.000 .1430003639040 3102 147 425000.000 .145000258 450000.000 .147000153 475000.000 .1490017709040 3102 148 500000.000 .151942351 525000.000 .155672651 550000.000 .1594029529040 3102 149 575000.000 .163122586 600000.000 .164940353 630000.000 .1650000009040 3102 150 660000.000 .165000000 690000.000 .164661912 720000.000 .1629724359040 3102 151 760000.000 .160945120 800000.000 .157571511 840000.000 .1527142539040 3102 152 880000.000 .147855060 920000.000 .142800070 960000.000 .1376001919040 3102 153 1000000.00 .130600244 1100000.00 .121800229 1200000.00 .1134002049040 3102 154 1300000.00 .107400117 1400000.00 .101800020 1500000.00 .0959999459040 3102 155 1600000.00 .089999894 1700000.00 .083749872 1800000.00 .0760000279040 3102 156 1900000.00 .068000212 2000000.00 .061800293 2100000.00 .0574002709040 3102 157 2200000.00 .053000247 2300000.00 .048600224 2400000.00 .0442002019040 3102 158 2500000.00 .040800183 2600000.00 .038400170 2700000.00 .0360001579040 3102 159 2800000.00 .033600144 2900000.00 .031200131 3000000.00 .0294001229040 3102 160 3100000.00 .028200118 3200000.00 .027000114 3300000.00 .0258001109040 3102 161 3400000.00 .024600106 3500000.00 .023400102 3600000.00 .0222000989040 3102 162 3700000.00 .021000094 3800000.00 .019800090 3900000.00 .0186000869040 3102 163 4000000.00 .017750083 4100000.00 .017250081 4200000.00 .0167500799040 3102 164 4300000.00 .016250077 4400000.00 .015750075 4500000.00 .0152500729040 3102 165 4600000.00 .014750070 4700000.00 .014250068 4800000.00 .0137500669040 3102 166 4900000.00 .013250064 5000000.00 .012750062 5100000.00 .0122500609040 3102 167 5200000.00 .011750058 5300000.00 .011250056 5400000.00 .0107500549040 3102 168 5500000.00 .010250051 5600000.00 .009750049 5700000.00 .0092500479040 3102 169 5800000.00 .008750045 5900000.00 .008250043 6000000.00 .0078950419040 3102 170 6100000.00 .007685040 6200000.00 .007475039 6300000.00 .0072650389040 3102 171 6400000.00 .007055037 6500000.00 .006845036 6600000.00 .0066350359040 3102 172 6700000.00 .006425034 6800000.00 .006215033 6900000.00 .0060050329040 3102 173 7000000.00 .005795030 7100000.00 .005585029 7200000.00 .0053750289040 3102 174 7300000.00 .005165027 7400000.00 .004955026 7500000.00 .0047450259040 3102 175 7600000.00 .004535024 7700000.00 .004325023 7800000.00 .0041150229040 3102 176 7900000.00 .003905021 8000000.00 .003755020 8100000.00 .0036650209040 3102 177 8200000.00 .003575019 8300000.00 .003485019 8400000.00 .0033950199040 3102 178 8500000.00 .003305019 8600000.00 .003215018 8700000.00 .0031250189040 3102 179 8800000.00 .003035018 8900000.00 .002945018 9000000.00 .0028550179040 3102 180 9100000.00 .002765017 9200000.00 .002675017 9300000.00 .0025850179040 3102 181 9400000.00 .002495016 9500000.00 .002405016 9600000.00 .0023150169040 3102 182 9700000.00 .002225016 9800000.00 .002135015 9900000.00 .0020450159040 3102 183 10000000.0 .001993765 10100000.0 .001981265 10200000.0 .0019687649040 3102 184 10300000.0 .001956264 10400000.0 .001943764 10500000.0 .0019312639040 3102 185 10600000.0 .001918763 10700000.0 .001906263 10800000.0 .0018937629040 3102 186 10900000.0 .001881262 11000000.0 .001868762 11100000.0 .0018562619040 3102 187 11200000.0 .001843761 11300000.0 .001831261 11400000.0 .0018187609040 3102 188 11500000.0 .001806260 11600000.0 .001793760 11700000.0 .0017812599040 3102 189 11800000.0 .001768759 11900000.0 .001756259 12000000.0 .0017437589040 3102 190 12100000.0 .001731258 12200000.0 .001718758 12300000.0 .0017062579040 3102 191 12400000.0 .001693757 12500000.0 .001681257 12600000.0 .0016687569040 3102 192 12700000.0 .001656256 12800000.0 .001643756 12900000.0 .0016312559040 3102 193 13000000.0 .001618755 13100000.0 .001606255 13200000.0 .0015937549040 3102 194 13300000.0 .001581254 13400000.0 .001568754 13500000.0 .0015562539040 3102 195 13600000.0 .001543753 13700000.0 .001531253 13800000.0 .0015187529040 3102 196 13900000.0 .001506252 14000000.0 .001499169 14100000.0 .0014975029040 3102 197 14200000.0 .001495835 14300000.0 .001494169 14400000.0 .0014925029040 3102 198 14500000.0 .001490835 14600000.0 .001489169 14700000.0 .0014875029040 3102 199 14800000.0 .001485835 14900000.0 .001484169 15000000.0 .0014825029040 3102 200 15100000.0 .001480835 15200000.0 .001479169 15300000.0 .0014775029040 3102 201 15400000.0 .001475835 15500000.0 .001474169 15600000.0 .0014725029040 3102 202 15700000.0 .001470835 15800000.0 .001469169 15900000.0 .0014675029040 3102 203 16000000.0 .001465835 16100000.0 .001464169 16200000.0 .0014625029040 3102 204 16300000.0 .001460835 16400000.0 .001459169 16500000.0 .0014575029040 3102 205 16600000.0 .001455835 16700000.0 .001454169 16800000.0 .0014525029040 3102 206 16900000.0 .001450835 17000000.0 .001449169 17100000.0 .0014475029040 3102 207 17200000.0 .001445835 17300000.0 .001444169 17400000.0 .0014425029040 3102 208 17500000.0 .001440835 17600000.0 .001439169 17700000.0 .0014375029040 3102 209 17800000.0 .001435835 17900000.0 .001434169 18000000.0 .0014325029040 3102 210 18100000.0 .001430835 18200000.0 .001429169 18300000.0 .0014275029040 3102 211 18400000.0 .001425835 18500000.0 .001424169 18600000.0 .0014225029040 3102 212 18700000.0 .001420835 18800000.0 .001419169 18900000.0 .0014175029040 3102 213 19000000.0 .001415835 19100000.0 .001414169 19200000.0 .0014125029040 3102 214 19300000.0 .001410835 19400000.0 .001409169 19500000.0 .0014075029040 3102 215 19600000.0 .001405835 19700000.0 .001404169 19800000.0 .0014025029040 3102 216 19900000.0 .001400835 20000000.0 0.0 9040 3102 217 9040 3 099999 9040 0 0 0 9.02320E+4 2.30044E+2 0 0 0 1904033 18 1 0.000000+0 0.000000+0 0 18 0 2904033 18 2 0.000000+0 0.000000+0 0 1 6 3904033 18 3 1.000000-5 0.000000+0 5.000000+5 1.600000-3 2.000000+7 0.000000+0904033 18 4 0.000000+0 0.000000+0 0 1 24 12904033 18 5 1.000000-5 0.000000+0 5.000000+5 1.024000-1 8.000000+5 1.440000-2904033 18 6 9.000000+5 1.000000-2 1.000000+6 4.900000-3 1.200000+6 3.600000-3904033 18 7 1.400000+6 1.600000-3 1.600000+6 3.600000-3 3.000000+6 1.600000-3904033 18 8 4.000000+6 3.600000-3 6.000000+6 1.000000-2 2.000000+7 0.000000+0904033 18 9 904033 099999 9.02320E+4 2.30044E+2 0 0 0 1904033102 1 0.000000+0 0.000000+0 0 102 0 2904033102 2 0.000000+0 0.000000+0 0 1 6 3904033102 3 1.000000-5 0.000000+0 1.500000+1 2.500000-3 2.000000+7 0.000000+0904033102 4 0.000000+0 0.000000+0 0 1 10 5904033102 5 1.000000-5 2.000000-4 3.000000-2 6.400000-3 1.500000+1 1.000000-2904033102 6 1.000000+6 4.000000-2 2.000000+7 0.000000+0 904033102 7 904033 099999 9040 0 0 0 0 0 0 0 9.22350E+4 2.33024E+2 0 1 34 109228 1451 1 0.0 1.00000E+0 0 0 0 69228 1451 2 1.00000E+0 2.00000E+7 0 0 10 20029228 1451 3 3.00000E+2 0.0 1 0 414 49228 1451 4 92-U -235 ORNL,LANL,+EVAL-NOV89 WESTON, YOUNG, POENITZ, LUBITZ 9228 1451 5 DIST-Feb2004 9228 1451 6 ----IRDF-2002 MATERIAL 9228 9228 1451 7 -----INCIDENT NEUTRON DATA 9228 1451 8 ------ENDF-6 FORMAT 9228 1451 9 **************************************************************** 9228 1451 10 92-U -235 ORNL,LANL,+EVAL-NOV89 WESTON, YOUNG, POENITZ, LUBITZ 9228 1451 11 DIST-NOV98 REV5-OCT97 19981109 9228 1451 12 ----ENDF/B-VI MATERIAL 9228 REVISION 5 9228 1451 13 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9228 1451 14 DOSIMETRY 9228 1451 15 ******************************************************************9228 1451 16 9228 1451 17 ENDF/B-VI MOD 6 Revision, October 1997, L.C. Leal, H. Derrien, 9228 1451 18 N.M. Larson, and R.Q. Wright (ORNL) 9228 1451 19 9228 1451 20 1. New resonance parameter analysis; File 2, MT=151. 9228 1451 21 2. Total and prompt-nubar revised below 2 eV; File 1, MT=451, 9228 1451 22 452,456. 9228 1451 23 9228 1451 24 ---------------------------------------------------------------- 9228 1451 25 File 2 9228 1451 26 MT=151 Resonance parameters, from a new analysis by Leal 9228 1451 27 et al. [LE97], using the multilevel R-matrix analysis code 9228 1451 28 SAMMY [LA96]. Energy range for U235 is 0 to 2.25 keV. 9228 1451 29 For the first time, integral data were fitted during the 9228 1451 30 analysis process: Thermal cross sections (fission, 9228 1451 31 capture, and elastic), Westcott g-factors (fission and 9228 1451 32 absorption) are from the ENDF/B-6 standards [CA93], and 9228 1451 33 the K1 value is from Hardy [HA79]. 9228 1451 34 Thermal parameters obtained in the present evaluation, 9228 1451 35 first using the microscopic experimental data only, and 9228 1451 36 second including the integral data as well, are compared 9228 1451 37 to the SAMMY input in the following Table: 9228 1451 38 9228 1451 39 Parameter SAMMY input Fit to Fit to diff. 9228 1451 40 value diff data & integ. 9228 1451 41 alone data 9228 1451 42 ----------- ----------------- -------- ------------ 9228 1451 43 Fission 584.25 +/- 1.11 582.28 584.88 9228 1451 44 Capture 98.96 +/- 0.74 99.18 98.66 9228 1451 45 Scattering 15.46 +/- 1.06 15.44 15.12 9228 1451 46 Westcott gf 0.9771 +/- 0.0008 0.9743 0.9764 9228 1451 47 Westcott ga 0.9790 +/- 0.0008 0.9774 0.9785 9228 1451 48 Westcott gg 0.9956 0.9910 9228 1451 49 K1(barn) 722.70 +/- 3.90 717.48 722.43 9228 1451 50 9228 1451 51 The final adjustment of nu by SAMMY to the recommended K1 9228 1451 52 value of 722.7 gave nu = 2.4367 +/- 0.0005, with fission 9228 1451 53 and absorption cross sections calculated from the final 9228 1451 54 resonance parameters. 9228 1451 55 In the following Tables, the fission and capture cross 9228 1451 56 sections calculated in this evaluation with the code 9228 1451 57 SAMMY are compared with experimental data. 9228 1451 58 9228 1451 59 Experimental and calculated total cross sections. 9228 1451 60 9228 1451 61 Energy Range Calculated Schrack Weston Weston 9228 1451 62 (eV) (b.eV) (b.eV) (b.eV) (b.eV) 9228 1451 63 --------------- ---------- ------- ------ ------ 9228 1451 64 0.5 - 20.0 910.4 929.9 9228 1451 65 20.0 - 60.0 1867.8 1882.8 1869.9 9228 1451 66 60.0 - 100.0 954.0 968.0 954.2 9228 1451 67 100.0 - 200.0 2032.7 2092.7 2089.5 2073.9 9228 1451 68 200.0 - 300.0 2062.2 2007.0 2060.0 2054.6 9228 1451 69 300.0 - 400.0 1280.8 1321.6 1297.1 1292.9 9228 1451 70 400.0 - 500.0 1333.2 1391.5 1351.8 1347.9 9228 1451 71 500.0 - 600.0 1489.2 1467.9 1499.2 1494.3 9228 1451 72 600.0 - 700.0 1126.6 1156.4 1134.1 1132.6 9228 1451 73 700.0 - 800.0 1088.7 1085.8 1093.3 1075.7 9228 1451 74 800.0 - 900.0 797.6 784.0 813.0 804.9 9228 1451 75 900.0 - 1000.0 724.4 723.9 738.2 721.4 9228 1451 76 1000.0 - 2000.0 7036.1 7054.2 9228 1451 77 9228 1451 78 Experimental and calculated capture cross sections. 9228 1451 79 9228 1451 80 Energy Range Calculated De Saussure Perez 9228 1451 81 (eV) (b.eV) (b.eV) (b.eV) 9228 1451 82 --------------- ---------- ----------- ------ 9228 1451 83 0.5 - 20.0 653.5 647 9228 1451 84 20.0 - 60.0 1066.1 1084 1057 9228 1451 85 60.0 - 100.0 490.2 477 504 9228 1451 86 100.0 - 200.0 1158.8 1148 1138 9228 1451 87 200.0 - 300.0 907.8 904 940 9228 1451 88 300.0 - 400.0 660.2 658 642 9228 1451 89 400.0 - 500.0 495.9 506 478 9228 1451 90 500.0 - 600.0 533.3 506 562 9228 1451 91 600.0 - 700.0 494.8 481 449 9228 1451 92 700.0 - 800.0 490.1 513 475 9228 1451 93 800.0 - 900.0 439.8 444 397 9228 1451 94 900.0 - 1000.0 504.2 542 482 9228 1451 95 1000.0 - 1100.0 509.6 522 463 9228 1451 96 1100.0 - 1200.0 413.7 395 332 9228 1451 97 1200.0 - 1300.0 340.4 372 267 9228 1451 98 1300.0 - 1400.0 304.1 304 225 9228 1451 99 1400.0 - 1500.0 355.7 301 254 9228 1451 100 --------------- ---------- ----------- ------ 9228 1451 101 20.0 - 1500.0 9164.7 9046 8665 9228 1451 102 9228 1451 103 The fission and capture resonance integral calculated from 9228 1451 104 the present evaluation are 276.04 b and 140.49 b, respectively, 9228 1451 105 giving a capture-to-fission ratio (alpha value) of 0.509 in 9228 1451 106 excellent agreement with the value obtained from integral 9228 1451 107 measurements. 9228 1451 108 The following energy-differential data were included in the 9228 1451 109 analysis: 9228 1451 110 (1) Transmission data of Harvey et al. [HA86] on the ORELA 9228 1451 111 18-meter flight path, with sample thickness of 0.03269 9228 1451 112 atoms/barn, cooled to 77 K (0.4 to 68 eV). 9228 1451 113 (2) Transmission data of Harvey et al. [HA86] on the ORELA 9228 1451 114 80-meter flight path, with sample thickness of 0.00233 9228 1451 115 atoms/barn, cooled to 77 K (4 to 2250 eV). 9228 1451 116 (3) Transmission data of Harvey et al. [HA86] on the ORELA 9228 1451 117 80-meter flight path, with sample thickness of 0.03269 9228 1451 118 atoms/barn, cooled to 77 K (4 to 2250 eV). 9228 1451 119 (4) Fission data of Schrack [SC88] on the RPI Linac at 8.4 9228 1451 120 meter flight path (0.02 to 20 eV). 9228 1451 121 (5,6) Fission and capture data of de Saussure et al. [DE67] 9228 1451 122 on the ORELA 25.2-meter flight path (0.01 to 2250 eV). 9228 1451 123 (7,8) Fission and capture data of Perez et al. [PE73] on the 9228 1451 124 ORELA 39-meter flight path (0.01 to 100 eV). 9228 1451 125 (9) Fission data of Gwin et al. [GW84] on the ORELA 25.6-meter 9228 1451 126 flight path (0.01 to 20 eV). 9228 1451 127 (10) Transmission data of Spencer et al. [SP84] on the ORELA 9228 1451 128 ORELA 18-meter flight path, sample thickness of 0.001468 9228 1451 129 atom/barn (0.01 to 1.0 eV). 9228 1451 130 (11) Fission data of Wagemans et al. [WA88] on the Geel 18- 9228 1451 131 meter flight path (0.001 to 1.0 eV) 9228 1451 132 (12,13) Absorption and fission data of Gwin [GW96] at ORELA 9228 1451 133 (0.01 to 4.0 eV). 9228 1451 134 (14) Fission data of Weston and Todd [WE84] on the ORELA 9228 1451 135 18.9-meter flight path (14 to 2250 eV). 9228 1451 136 (15) Eta data of Wartena et al. [WA87] at 8 meters (0.0018 to 9228 1451 137 1.0 eV). 9228 1451 138 (16) Eta (chopper) data of Weigmann et al [WE90] (0.0015 to 9228 1451 139 0.15 eV). 9228 1451 140 (17) Fission data of Weston and Todd [WE92] on the ORELA 9228 1451 141 86.5-meter flight path (100 to 2000 eV). 9228 1451 142 (18) Fission yield data of Moxon et al. [MO92] at ORELA 9228 1451 143 (0.01 to 50.0 eV). 9228 1451 144 9228 1451 145 ---------------------------------------------------------------- 9228 1451 146 REFERENCES FOR RESOLVED RESONANCE REGION 9228 1451 147 9228 1451 148 [CA93] A. Carlson, W.P. Poenitz, G.M. Hale et al., "The ENDF/B-6 9228 1451 149 Neutron Cross Section Measurements Standards," National 9228 1451 150 Institute of Standards and Technology report NISTIR-5177 9228 1451 151 (1993) 9228 1451 152 [DE67] G. de Saussure, R. Gwin, L.W. Weston, and R.W. Ingle, 9228 1451 153 "Simultaneous Measurements of the Neutron Fission and Capture 9228 1451 154 Cross Section for 235U for Incident Neutron Energy from 9228 1451 155 0.04 eV to 3 keV," Oak Ridge National Laboratory report 9228 1451 156 ORNL/TM-1804 (1967) 9228 1451 157 [GW84] R. Gwin, R.R. Spencer, R.W. Ingle, J.H. Todd, and S.W. 9228 1451 158 Scoles, Nuc.Sci.Eng. 88, 37 (1984) 9228 1451 159 [GW96] R. Gwin, To be published in Nuclear Science Engineering 9228 1451 160 [HA79] J. Hardy, Brookhaven National Laboratory, report 9228 1451 161 BNL-NCS-51123 [ENDF-300] (1979) Sec. B.1 9228 1451 162 [HA86] J.A. Harvey, N.W. Hill, F.G. Perey et al., Nuclear Data 9228 1451 163 for Science and Technology, Proc. Int. Conf. May 30-June 3, 9228 1451 164 1988, Mito, Japan. (Saikon Publishing, 1988) p. 115 9228 1451 165 [LA96] N.M. Larson, "Updated Users' Guide to SAMMY" report 9228 1451 166 ORNL/TM-9179/R3 (1996) 9228 1451 167 [LE97] L.C. Leal, H. Derrien, N.M. Larson, R.Q. Wright, 9228 1451 168 "R-Matrix Analysis of 235U Neutron Transmission and Cross 9228 1451 169 Sections in the Energy Range 0 eV to 2.25 keV," Oak Ridge 9228 1451 170 National Laboratory report ORNL/TM-13516 (1997). 9228 1451 171 [MO92] M.C. Moxon, J.A. Harvey, and N.W. Hill, private 9228 1451 172 communication, Oak Ridge National Laboratory (1992). 9228 1451 173 [PE73] R.B. Perez, G. de Saussure, and E.G. Silver, Nucl.Sci. 9228 1451 174 Eng. 52, 46 (1973) 9228 1451 175 [SC88] R.A. Schrack, "Measurement of the 235U(n,f) Reaction from 9228 1451 176 Thermal to 1 keV," Nuclear Data for Science and Technology, 9228 1451 177 Proc. Int. Conf. May 30-June 3, Mito, Japan (Saikon 9228 1451 178 Publishing, 1988) p. 101 9228 1451 179 [SP84] R.R. Spencer, J.A. Harvey, N.W. Hill, and L. Weston, 9228 1451 180 Nucl.Sci.Eng. 96, 318 (1987) 9228 1451 181 [WA87] J.A. Wartena, H. Weigmann, and C. Burkholz, report IAEA 9228 1451 182 Tecdoc 491 (1987) p.123 9228 1451 183 [WA88] C. Wagemans, P. Schillebeeckx, A.J. Deruyter, and R. 9228 1451 184 Barthelemy, "Subthermal Fission Cross Section Measurements 9228 1451 185 for 233U and 239Pu," Nuclear Data for Science and Technology, 9228 1451 186 Proc. Int. Conf. May 30-June 3, Mito, Japan (Saikon 9228 1451 187 Publishing, 1988) p. 91 9228 1451 188 [WE84] L.W. Weston and J.H. Todd, Nucl.Sci.Eng. 88, 567 (1984) 9228 1451 189 [WE90] H. Weigmann, P. Geltenbort, B. Keck, K. Shrenckenbach, 9228 1451 190 and J.A. Wartena, The Physics of Reactors, Proc. Int. Conf., 9228 1451 191 Marseille, 1990, Vol.1 (1990) p. 133 9228 1451 192 [WE92] L.W. Weston and J.H. Todd, Nucl.Sci.Eng. 111, 415 (1992) 9228 1451 193 9228 1451 194 **************************************************************** 9228 1451 195 9228 1451 196 ENDF/B-VI MOD 5 Revision 4 (R.Q.Wright, ORNL, V. McLane, NNDC, 9228 1451 197 October, 1996) 9228 1451 198 9228 1451 199 File 1, MT=451: Update of comments. 9228 1451 200 File 1, MT=456: Update of prompt nubar. 9228 1451 201 9228 1451 202 *****************************************************************9228 1451 203 9228 1451 204 ENDF/B-VI MOD 4 Revision, February 1995, C. Lubitz (KAPL) 9228 1451 205 9228 1451 206 The resonance parameters below 900 eV were adjusted as described 9228 1451 207 in C.R. Lubitz, "A modification to ENDF6 U235 to increase 9228 1451 208 epithermal alpha and K1", Proceedings of the Int. Conf. on 9228 1451 209 Nuclear Data for Science and Technology, Gatlinburg, TN, May 9228 1451 210 9-13, Vol. 2 (American Nuclear Soc., 1994) page 646. 9228 1451 211 9228 1451 212 *****************************************************************9228 1451 213 9228 1451 214 ENDF/B-VI MOD 3 Revision 2 9228 1451 215 9228 1451 216 Missing File 1 section 458 added. 9228 1451 217 9228 1451 218 *****************************************************************9228 1451 219 9228 1451 220 ENDF/B-VI MOD 2 Revision, L. Weston (ORNL) 9228 1451 221 9228 1451 222 FILE 1: The uncertainties on the fission cross section 9228 1451 223 recommended by the Standards Committee of CSEWG are listed. The 9228 1451 224 updated description of File 2, the resolved resonance parameters,9228 1451 225 is given. 9228 1451 226 9228 1451 227 FILE 2: The resolved resonance parameters have been changed 9228 1451 228 extensively. The neutron energy region below 4 eV has been made 9228 1451 229 a separate group. Eta decreases with decreasing neutron energy 9228 1451 230 below 0.1 eV. At the higher neutron energies, the resonance 9228 1451 231 parameters are more refined. 9228 1451 232 9228 1451 233 FILE 3: Smooth cross sections above 100 keV have minor 9228 1451 234 corrections. 9228 1451 235 9228 1451 236 FILE 31: Nubar covariance files were updated. 9228 1451 237 9228 1451 238 FILE 33: All covariance files were removed as correct new files 9228 1451 239 are not yet available. 9228 1451 240 *****************************************************************9228 1451 241 9228 1451 242 ENDF/B-VI, MOD 1 Evaluation, April 1989 9228 1451 243 9228 1451 244 Principal evaluators: 9228 1451 245 Thermal parameters: Standards Committee of CSEWG. 9228 1451 246 Resolved resonance region: (0-2250 eV) L.C. Leal (U.Tenn.), 9228 1451 247 G. deSaussure (ORNL), R.B. Perez (U.Tenn.), N.M. Larson 9228 1451 248 (ORNL), M.S.Moore (LANL), and R.Q. Wright (ORNL). 9228 1451 249 Replaced by revision 6. 9228 1451 250 Unresolved resonance region and File 3 below 100 keV and the 9228 1451 251 capture cross section above 100 keV: W.P. Poenitz (ANL) and 9228 1451 252 L.W. Weston (ORNL) 9228 1451 253 Fission cross section above 100 keV: W.P. Poenitz (ANL) and the 9228 1451 254 Standards Committee of CSEWG. 9228 1451 255 Model calculations and fits above 100 keV: P.G. Young (LANL), 9228 1451 256 R.E. MacFarlane (LANL), and E.D. Arthur (LANL). 9228 1451 257 Covariance files: R.W. Peelle (ORNL). All files reoved in 9228 1451 258 revision 1, except MT=456. 9228 1451 259 9228 1451 260 File 1 Descriptive and Nubar Information ----------------------- 9228 1451 261 MT=452 Total nubar. Replaced in revision 6. 9228 1451 262 MT=455 Delayed neutron yields from England [EN89]. 9228 1451 263 MT=456 Prompt neutron yields. Replaced in revision 6. 9228 1451 264 9228 1451 265 FISSION CROSS SECTION UNCERTAINTIES: The CSEWG Standards 9228 1451 266 Committee supplied the following uncertainties with the these 9228 1451 267 statements. "These uncertainties are estimates such that if a 9228 1451 268 modern day experiment were performed today on a given standard 9228 1451 269 using the best techniques, those results should fall within these9228 1451 270 expanded uncertainties (2/3 of the time). They take into account9228 1451 271 inconsistencies and concerns about R-matrix parameters. Note 9228 1451 272 that it is not assumed that the uncertainties are totally 9228 1451 273 correlated within the energy ranges given." 9228 1451 274 9228 1451 275 Energy (keV) Estimated Comb Result (%) 9228 1451 276 Uncertainty (%) 9228 1451 277 2.53E-05 0.2 (0.19) 9228 1451 278 150-600 1.5 9228 1451 279 600-1000 1.6 (0.60) 9228 1451 280 1000-3000 1.8 9228 1451 281 3000-6000 2.3 (0.69) 9228 1451 282 6000-10000 2.2 9228 1451 283 10000-12000 1.8 (1.14) 9228 1451 284 12000-14000 1.2 9228 1451 285 14000-14500 0.8 (0.55) 9228 1451 286 14500-15000 1.5 9228 1451 287 15000-16000 2.0 (0.97) 9228 1451 288 16000-17000 2.5 9228 1451 289 17000-19000 3.0 (1.26) 9228 1451 290 19000-20000 4.0 9228 1451 291 9228 1451 292 File 2. RESOLVED AND UNRESOLVED RESONANCE PARAMETERS ---------- 9228 1451 293 MT=151 9228 1451 294 Resolved resonance region: replaced in revision 6. 9228 1451 295 Unresolved resonance region: 9228 1451 296 The unresolved resonance region was derived by a FITACS 9228 1451 297 (Fritz Froehner code) fit by L.W. Weston to the Standards 9228 1451 298 Committee recommendation for the fission cross section and 9228 1451 299 new capture evaluation based on newer alpha measurements 9228 1451 300 (see ANL-83-4 supplement). These results were then fit 9228 1451 301 with URES (Ed Pennington code) so ENDF would reproduce. 9228 1451 302 unresolved resonance region extends from 2.25 to 25 keV 9228 1451 303 and is used only for self shielding calculations. Dilute 9228 1451 304 cross sections are taken from File 3 which shows 9228 1451 305 experimentally observed structure carried over from 9228 1451 306 version 5 up to 100 keV. 9228 1451 307 9228 1451 308 File 3. SMOOTH CROSS SECTIONS ---------------------------------- 9228 1451 309 Model calculations from LANL: P.G. Young, R.E.MacFarlane, 9228 1451 310 E.D.Arthur. The evaluation above 100 keV is based on a detailed 9228 1451 311 theoretical analysis utilizing the available experimental data. 9228 1451 312 Coupled channel optical model calculations with the ECIS code 9228 1451 313 [Ra70] were used to provide the total, elastic, and inelastic 9228 1451 314 cross sections to the first 3 members of the ground state 9228 1451 315 rotational band, as well as neutron elastic and inelastic 9228 1451 316 angular distributions to the rotational levels. The ECIS code 9228 1451 317 was also used to calculate neutron transmission coefficients. 9228 1451 318 Hauser-Feshbach statistical theory calculations were carried out 9228 1451 319 with the GNASH [Ar88,Yo77] and COMNUC [Du70] code systems, 9228 1451 320 including preequilibrium and fission. DWBA calculations were 9228 1451 321 performed with the DWUCK code [Ku70] for several vibrational 9228 1451 322 levels, using B(El) values inferred from (d,d') data on U234, 9228 1451 323 U235, U238, as well as Coulomb excitation measurements. A weak 9228 1451 324 coupling model [Pe69] was used to apply the U234 and U238 9228 1451 325 results to states in U235. 9228 1451 326 A preliminary description of the analysis was given at the 9228 1451 327 Mito conference [Yo88]. 9228 1451 328 9228 1451 329 MT=1 SUM OF PARTIAL CROSS SECTIONS FROM 2.25 TO 100 KEV. 9228 1451 330 0.10 to 20 Mev, obtained from a covariance analysis of 9228 1451 331 available experimental data, using an initial or prior cross 9228 1451 332 section from the coupled-channel optical model analysis. 9228 1451 333 Experimental data used include [Fo71], [Ve80], [Bo72], [Po81], 9228 1451 334 [Gr73], [Sc74], [Po83], [Pe60], [Wh65], [Ca73], and [Br58]. 9228 1451 335 The GLUCS code was used for analysis [He80]. 9228 1451 336 MT=2 UNCHANGED FROM VERSION 5 FROM 2.25 TO 100 KEV. 9228 1451 337 0.12 to 20 MeV, based on subtraction of (MT=4,16,17,18,37,102) 9228 1451 338 from MT=1. 9228 1451 339 MT=18 2.2 to 120 keV: average values from simultaneous 9228 1451 340 evaluation of W.P. Poenitz with structure carried over from 9228 1451 341 version 5. Renormalizaion was over 3 ranges per decade. 9228 1451 342 0.10 to 20 MeV: standards evaluation by CSEWG Standards 9228 1451 343 Committee. 9228 1451 344 9228 1451 345 ---------------------------------------------------------------- 9228 1451 346 REFERENCES 9228 1451 347 9228 1451 348 [Ar88] E.D. Arthur, Los Alamos National Laboratory report 9228 1451 349 LA-UR-88-382 (1988) 9228 1451 350 [Bo72] K. Boeckhoff et al., J.Nuc.En. 26, 91 (1972) 9228 1451 351 [Br58] A. Bratenahl et al., Phys.Rev. 110, 927 (1958) 9228 1451 352 [Ca73] J. Cabe et al., report CEA-R-4524 (1973) 9228 1451 353 [Du70] C.L. Dunford, report AI-AEC-12931 (1970) 9228 1451 354 [En89] T.R. England et al, Los Alamos National Laboratory 9228 1451 355 reports LA-11151-MS(88), LA-11534T(89); LAUR-88-4118 to be 9228 1451 356 published in Nucl.Sci.Eng. (1989) 9228 1451 357 [Fo71] D. Foster and D. Glasgow, Phys.Rev. C 3, 576 (1971) 9228 1451 358 [Fr86] J. Frehaut, report NEANDC(E) 238/L (1986) 9228 1451 359 [Gr73] L. Green et al., report USNDC-9 (1973) p.170 9228 1451 360 [He80] D. Hetrick and C.Y. Fu, Oak Ridge National Laboratory 9228 1451 361 report ORNL/TM-7341 (1980) 9228 1451 362 [Ku70] P.D. Kunz, "DWUCK: A Distorted-Wave Born Approximation 9228 1451 363 Program," unpublished report 9228 1451 364 [Ma88] D.G.Madland, Nuclear Data for Science and Technology, 9228 1451 365 Proc. Int. Conf., Mito, Japan, May 30-June 3, 1988, Mito, 9228 1451 366 Japan. (Saikon Publishing, 1988) p.759 9228 1451 367 [Pe60] J. Peterson et al., Phys.Rev. 120, 521 (1960) 9228 1451 368 [Pe69] R.J. Peterson, Ann.Phys. 53, 40 (1069) 9228 1451 369 [Po81] W. Poenitz et al., Nuc.Sci.Eng. 78, 333 (1981) 9228 1451 370 [Po83] W. Poenitz et al., Argonne National Laboratory report 9228 1451 371 ANL-NDM-80 (1983) 9228 1451 372 [Ra70] J. Raynal, report IAEA SMR-9/8 (1972) p.281 9228 1451 373 [Sc74] R. Schwartz et al., Nuc.Sci.Eng. 54, 322 (1974) 9228 1451 374 [Ut66] C. Uttley et al., Paris Conf. (1966) v1, p165 9228 1451 375 [Ve80] V. Vertebnyj et al., report YFI-16,8(1973) 9228 1451 376 [Wh65] W. Whalen et al., Argonne National Laboratory report 9228 1451 377 ANL-7110 (1965) p.15 9228 1451 378 [Yo77] P.G. Young and E.D. Arthur, Los Alamos National 9228 1451 379 Laboratory report LA-6947 (1977). 9228 1451 380 [Yo88] P.G. Young and E.D. Arthur, Nuclear Data for Science and 9228 1451 381 Technology, Proc. Int. Conf., Mito, Japan, May 30-June 3, 9228 1451 382 1988, Mito, Japan. (Saikon Publishing, 1988) p.603 9228 1451 383 9228 1451 384 *****************************************************************9228 1451 385 9228 1451 386 9228 1451 387 9228 1451 388 9228 1451 389 ************************ C O N T E N T S *********************** 9228 1451 390 ***************** Program LINEAR (VERSION 2002-1) ***************9228 1451 391 For All Data Greater than 1.0000E-10 barns in Absolute Value 9228 1451 392 Data Linearized to Within an Accuracy of .100000000 per-cent 9228 1451 393 ***************** Program RECENT (VERSION 2002-1) ***************9228 1451 394 for All Data Greater than 1.0000E-10 barns in Absolute Value 9228 1451 395 Data Linearized to within an Accuracy of .100000000 per-cent 9228 1451 396 ***************** Program SIGMA1 (VERSION 2002-1) ***************9228 1451 397 Data Doppler Broadened to 300.000000 Kelvin 9228 1451 398 for All Data Greater than 1.0000E-10 barns in Absolute Value 9228 1451 399 Data Linearized to Within an Accuracy pf .100000000 per-cent 9228 1451 400 ***************** Program FIXUP (Version 2002-1) ****************9228 1451 401 Corrected ZA/AWR in All Sections-----------------------------Yes 9228 1451 402 Corrected Thresholds-----------------------------------------Yes 9228 1451 403 Extended Cross Sections to 20 MeV----------------------------No 9228 1451 404 Allow Cross Section Deletion---------------------------------No 9228 1451 405 Allow Cross Section Reconstruction---------------------------No 9228 1451 406 Make All Cross Sections Non-Negative-------------------------Yes 9228 1451 407 Delete Energies Not in Ascending Order-----------------------Yes 9228 1451 408 Deleted Duplicate Points-------------------------------------Yes 9228 1451 409 Check for Ascending MAT/MF/MT Order--------------------------Yes 9228 1451 410 Check for Legal MF/MT Numbers--------------------------------Yes 9228 1451 411 Allow Creation of Missing Sections---------------------------No 9228 1451 412 Allow Insertion of Energy Points-----------------------------No 9228 1451 413 Create Uniform Energy Grid-----------------------------------No 9228 1451 414 Delete Section if Cross Section =0 at All Energies-----------Yes 9228 1451 415 ***************** Program GROUPIE (VERSION 2002-1) **************9228 1451 416 Unshielded Group Averages Using 640 Groups 9228 1451 417 Weighting Spectrum: Flat (Constant) Spectrum 9228 1451 418 1 451 422 69228 1451 419 2 151 4 69228 1451 420 3 18 217 29228 1451 421 33 18 53 19228 1451 422 9228 1 099999 9228 0 0 0 9.22350E+4 2.33024E+2 0 0 1 09228 2151 1 9.22350E+4 1.00000E+0 0 0 1 09228 2151 2 1.00000E-5 2.25000E+3 0 0 0 09228 2151 3 3.50000E+0 9.60200E-1 0 0 0 09228 2151 4 9228 2 099999 9228 0 0 0 9.22350E+4 2.33024E+2 0 0 0 09228 3 18 1 1.93720E+8 1.93720E+8 0 0 1 6419228 3 18 2 641 1 9228 3 18 3 .000100000 9805.44565 .000105000 9576.58959 .000110000 9359.172989228 3 18 4 .000115000 9158.81971 .000120000 8924.60065 .000127500 8664.732699228 3 18 5 .000135000 8427.28509 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9.500000+5 1.325000+6 1.900000+6 2.900000+6922833 18 6 4.600000+6 6.750000+6 9.500000+6 1.350000+7 1.550000+7 2.000000+7922833 18 7 3.634398-6 3.634398-6 3.101778-6 3.101778-6 2.475285-6 2.322641-6922833 18 8 1.959580-6 1.703434-6 1.466791-6 1.351542-6 1.325030-6 1.266146-6922833 18 9 1.194608-6 1.079757-6 9.347829-7 9.056430-7 8.066544-7 7.272780-7922833 18 10 7.101887-7 6.671650-7 5.787847-7 3.145285-7 5.273911-7 4.634398-6922833 18 11 3.101778-6 3.101778-6 2.475285-6 2.322641-6 1.959580-6 1.703434-6922833 18 12 1.466791-6 1.351542-6 1.325030-6 1.266146-6 1.194608-6 1.079757-6922833 18 13 9.347829-7 9.056430-7 8.066544-7 7.272780-7 7.101887-7 6.671650-7922833 18 14 5.787847-7 3.145285-7 5.273911-7 1.789157-5 1.789157-5 7.247852-6922833 18 15 6.727942-6 5.540369-6 4.708277-6 3.992362-6 3.642562-6 3.577145-6922833 18 16 3.415922-6 3.207211-6 2.857612-6 2.437250-6 2.338613-6 2.053255-6922833 18 17 1.758301-6 1.688479-6 1.584746-6 1.370763-6 7.259742-7 1.252452-6922833 18 18 1.889157-5 7.247852-6 6.727942-6 5.540369-6 4.708277-6 3.992362-6922833 18 19 3.642562-6 3.577145-6 3.415922-6 3.207211-6 2.857612-6 2.437250-6922833 18 20 2.338613-6 2.053255-6 1.758301-6 1.688479-6 1.584746-6 1.370763-6922833 18 21 7.259742-7 1.252452-6 1.515003-5 1.175065-5 9.613425-6 8.353013-6922833 18 22 7.031353-6 6.360433-6 6.235565-6 5.958995-6 5.586026-6 4.961029-6922833 18 23 4.215194-6 4.034481-6 3.527568-6 2.977505-6 2.845209-6 2.669809-6922833 18 24 2.307214-6 1.212660-6 2.109852-6 1.425651-5 1.027029-5 9.264273-6922833 18 25 8.092281-6 7.193645-6 6.909848-6 6.597376-6 6.189397-6 5.529027-6922833 18 26 4.687975-6 4.474730-6 3.910263-6 3.294562-6 3.145358-6 2.949774-6922833 18 27 2.547051-6 1.336222-6 2.329611-6 1.828855-5 1.084003-5 1.010942-5922833 18 28 8.788468-6 8.118550-6 7.624354-6 7.217124-6 6.596855-6 5.637136-6922833 18 29 5.393782-6 4.707867-6 3.960688-6 3.778518-6 3.541295-6 3.057533-6922833 18 30 1.601948-6 2.796258-6 1.713836-5 1.346428-5 1.135534-5 9.803183-6922833 18 31 8.931132-6 8.471377-6 7.776282-6 6.667301-6 6.315108-6 5.495624-6922833 18 32 4.604988-6 4.386879-6 4.109636-6 3.535872-6 1.845778-6 3.233329-6922833 18 33 2.472297-5 1.420676-5 1.121323-5 1.012406-5 9.380244-6 8.650187-6922833 18 34 7.491573-6 7.060471-6 6.124241-6 5.104539-6 4.857984-6 4.560243-6922833 18 35 3.917419-6 2.041280-6 3.581805-6 2.452771-5 1.363176-5 1.080702-5922833 18 36 9.786191-6 8.890658-6 7.810343-6 7.390222-6 6.436816-6 5.432394-6922833 18 37 5.172414-6 4.876118-6 4.149269-6 2.158182-6 3.787016-6 2.330751-5922833 18 38 1.243220-5 1.034842-5 9.414536-6 8.300378-6 7.862886-6 6.840450-6922833 18 39 5.730159-6 5.447138-6 5.150638-6 4.365020-6 2.258554-6 3.983342-6922833 18 40 2.215062-5 1.231017-5 9.937610-6 8.819333-6 8.360300-6 7.232673-6922833 18 41 5.890169-6 5.557986-6 5.194965-6 4.487461-6 2.331275-6 4.112111-6922833 18 42 2.101510-5 1.116621-5 9.236808-6 8.618871-6 7.453539-6 6.004822-6922833 18 43 5.674326-6 5.321985-6 4.594350-6 2.389630-6 4.213657-6 1.824382-5922833 18 44 1.102998-5 9.652663-6 8.263210-6 6.798812-6 6.440576-6 6.060138-6922833 18 45 5.216436-6 2.681023-6 4.772513-6 1.857572-5 1.154913-5 9.302521-6922833 18 46 7.887074-6 7.561684-6 7.140802-6 6.174039-6 3.135542-6 5.679011-6922833 18 47 1.951436-5 1.048593-5 8.568757-6 8.226505-6 7.723123-6 6.680752-6922833 18 48 3.392445-6 6.159440-6 1.662942-5 1.010621-5 9.422602-6 9.006112-6922833 18 49 7.813030-6 4.480511-6 7.256347-6 2.313803-5 1.605371-5 1.369189-5922833 18 50 1.141800-5 5.553546-6 9.669227-6 3.160813-5 1.991877-5 1.359895-5922833 18 51 6.513847-6 1.172085-5 5.091886-5 2.121602-5 9.574444-6 1.525340-5922833 18 52 5.533352-5 1.220843-5 1.470634-5 2.737248-5 1.530720-5 8.062025-5922833 18 53 922833 099999 9228 0 0 0 0 0 0 0 9.22380E+4 2.36005E+2 0 1 34 49237 1451 1 0.0 1.00000E+0 0 0 0 69237 1451 2 1.00000E+0 2.00000E+7 0 0 10 20029237 1451 3 3.00000E+2 0.0 1 0 423 69237 1451 4 92-U -238 ORNL+,JAER EVAL-NOV89 L.W.WESTON ET AL.,Y.KANDA ET AL. 9237 1451 5 DIST-OCT98 REV3-FEB97 19981007 9237 1451 6 ----IRDF-2002 MATERIAL 9237 REVISION 3 9237 1451 7 -----INCIDENT NEUTRON DATA 9237 1451 8 ------ENDF-6 FORMAT 9237 1451 9 Relative covariances MF33 MT102 in the energy range up to 100 eV 9237 1451 10 were increased to account for the systematic difference between 9237 1451 11 the present ENDF/B-VI value of 2.718 barns and the newly 9237 1451 12 recommended value of 2.683 barns for the thermal capture cross 9237 1451 13 section. The correction amounts to ((2.718-2.683)/2.683)**2. 9237 1451 14 Ref.: A.Trkov et al: Re-visiting the U-238 Thermal Capture Cross 9237 1451 15 Section and Gamma-ray Emission Probabilities of Np-239 Decay, 9237 1451 16 to be published in Nuclear Science and Engineering, 2005. 9237 1451 17 ******************************************************************9237 1451 18 92-U -238 ORNL,LANL+ EVAL-NOV89 L.W.WESTON,P.G.YOUNG,W.POENITZ 9237 1451 19 DIST-OCT98 REV3-FEB97 19981007 9237 1451 20 ----ENDF/B-VI MATERIAL 9237 REVISION 3 9237 1451 21 *****************EXTRACT FOR SPECIAL PURPOSE FILE*****************9237 1451 22 DOSIMETRY 9237 1451 23 ******************************************************************9237 1451 24 92-U -238 KYU,JAERI+ EVAL-MAR94 Y.KANDA ET AL. 9237 1451 25 DIST-JUL98 9237 1451 26 ----JENDL/D-99 MATERIAL 9237 9237 1451 27 9237 1451 28 9237 1451 29 U -238 FISSION 9237 1451 30 **************************************************************** 9237 1451 31 9237 1451 32 ENDF/B-VI MOD 4 Revision, February 1997, V.McLane (NNDC) 9237 1451 33 9237 1451 34 MT=1 changes 9237 1451 35 MF=451: Updated comments, references. 9237 1451 36 MF=455: Corrected TAB1 bookkeeping (minor revision). 9237 1451 37 MT=2 changes 9237 1451 38 MF=151 9237 1451 39 Added l-dependent scattering radius equal to scattering 9237 1451 40 radius for resolved region. 9237 1451 41 Corrected energy upper limit in unresolved region. 9237 1451 42 9237 1451 43 **************************************************************** 9237 1451 44 9237 1451 45 ENDF/B-VI MOD 3 Revision, January 1993, L. Weston (ORNL) 9237 1451 46 9237 1451 47 MF=2 changes 9237 1451 48 Resolved region (to 10 keV): Replaces the preliminary 9237 1451 49 evaluation of Moxon and Sowerby with the final version 9237 1451 50 [Mo??]. 9237 1451 51 Unresolved region: Uses Froehner's evaluation. 9237 1451 52 9237 1451 53 **************************************************************** 9237 1451 54 9237 1451 55 ENDF/B-VI MOD 2 Revision 1, January 1991 (NNDC) 9237 1451 56 9237 1451 57 Covariance files for total, elastic scattering, fission and 9237 1451 58 capture cross sections removed, since the correct files were and 9237 1451 59 still are not yet available. 9237 1451 60 9237 1451 61 **************************************************************** 9237 1451 62 9237 1451 63 ENDF/B-VI MOD 1 Evaluation, November 1989 (ORNL, LANL, ANL) 9237 1451 64 9237 1451 65 Principal evaluators 9237 1451 66 -------------------- 9237 1451 67 RESOLVED RESONANCE REGION: (10-5 to 10,000 eV) 9237 1451 68 M.G.Sowerby and M. C. Moxon (Harwell) 9237 1451 69 UNRESOLVED RESONANCE REGION: (10 TO 149 keV) 9237 1451 70 F.H.Froehner(KFK) and W.P.Poenitz(ANL) 9237 1451 71 REACTIONS ABOVE 149 keV: 9237 1451 72 MODEL CODE CALCULATIONS: (LANL)-P.G.Young and R.E.MacFarlane 9237 1451 73 FISSION: W.P.Poenitz(ANL) 9237 1451 74 NU-BAR DELAYED AND DELAYED NEUTRON SPECTRA: 9237 1451 75 KAISER AND CARPENTER (Ka78) 9237 1451 76 GAMMA PRODUCTION FILES: R. HOWERTON (LLL) 9237 1451 77 UNCERTAINTY FILES: NOT YET AVAILABLE 9237 1451 78 9237 1451 79 ---------------------------------------------------------------- 9237 1451 80 9237 1451 81 ENDF/B-VI EVALUATION ABOVE THE UNRESOLVED RESONANCE REGION 9237 1451 82 P.G.Young and R.E.MacFarlane (LANL) 9237 1451 83 Updated by R.Q. Wright (ORNL) 9237 1451 84 9237 1451 85 Above the unresolved resonance region, new evaluations were 9237 1451 86 performed of the neutron total, (n,2n), (,n,3n), (n,4n), (n,f), 9237 1451 87 (n,nf), (n,2nf), (n,3nf), and (n,gamma) cross sections as well as9237 1451 88 prompt nubar. The elastic and inelastic data from ENDF/B-V were 9237 1451 89 carried over for Version VI. 9237 1451 90 To provide the new data, coupled channel optical model calcu- 9237 1451 91 lations were performed with the ECIS code (Ra70) for the lowest 9237 1451 92 3 members of the U238 ground state rotational band. These calcu- 9237 1451 93 lations were used to provide initial (prior) values for a covar- 9237 1451 94 iance analysis of the total cross section and to provide neutron 9237 1451 95 transmission coefficients for nuclear reaction theory 9237 1451 96 calculations with the GNASH (Ar88,Yo77) and COMNUC (Du70) Hauser-9237 1451 97 Feshbach statistical/fission/preequilibrium codes. These theory 9237 1451 98 calculations were used to provide the MF=6 neutron distributions 9237 1451 99 from the (n,2n), (n,3n), and (n,4n) reactions as well as prior 9237 1451 100 values for covariance analyses of the cross sections for those 9237 1451 101 reactions. Additionally, the above analyses plus DWBA calcu- 9237 1451 102 lations were used to check the ENDF/B-V evaluation of elastic 9237 1451 103 and inelastic scattering. While some differences were found, 9237 1451 104 the earlier work was generally found to be reliable, and we 9237 1451 105 decided to carry over the ENDF/B-V data because of the effort 9237 1451 106 taken to match experimental data, both at lower energies and at 9237 1451 107 14 MeV. 9237 1451 108 9237 1451 109 MF=1 DESCRIPTIVE INFORMATION 9237 1451 110 9237 1451 111 MF=2 RESONANCE PARAMETERS (Updated for MOD 3) 9237 1451 112 MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETERS 9237 1451 113 RESOLVED RESONANCE PARAMETERS (1.0E-5 eV to 10.0 keV) 9237 1451 114 Reich-Moore analysis from new evaluation by Moxon and 9237 1451 115 Sowerby [1]. Present parameters are a preliminary set 9237 1451 116 due to use of preliminary capture yield data of 9237 1451 117 Macklin et al. [2]. Parameters above 4.0 kev corrected9237 1451 118 for background and normalisation errors in [2]. 9237 1451 119 Parameters 0f the bound resonances evaluated so as to 9237 1451 120 reproduce the following values at 2200m/s 9237 1451 121 9237 1451 122 Reaction Cross section (barns) Reference 9237 1451 123 Elastic 9.38 [3] 9237 1451 124 Fission 11.0E-6 [4] 9237 1451 125 Capture 2.708 [5],[6] 9237 1451 126 RESOLVED RESONANCE RANGE (10.0 to 149 keV) 9237 1451 127 Coherent calculation of point cross sections and of 9237 1451 128 energy-dependent average resonance parameters from 9237 1451 129 evaluated average resonance parameters for E = 0 with 9237 1451 130 Hauser-Feshbach code FITACS [7] 9237 1451 131 In the unresolved region the point cross sections given 9237 1451 132 in MF=3 are considered as correct. The unresolved 9237 1451 133 parameters in MF=2 are only to be used for self- 9237 1451 134 shielding calculations. The use of the parameters with 9237 1451 135 ENDF processing codes does not lead to cross sections 9237 1451 136 consistent with MF=3 as the codes use a more primative 9237 1451 137 version of the formalism. 9237 1451 138 Cross-sections used to determine the parameters 9237 1451 139 TOTAL: determined mainly by the following data. 9237 1451 140 below 45 keV: Uttley et al. [19], Byoun et al. [20], 9237 1451 141 Kononov and Poletaev [9], Tsubone et al. [8], 9237 1451 142 above 45 keV: Smith et al. [10] 9237 1451 143 ELASTIC SCATTERING: 9237 1451 144 Calculated as (total)-(partial cross-sections) 9237 1451 145 TOTAL INELASTIC AND INELASTIC LEVELS MT=51 AND 52 9237 1451 146 Determined mainly by the following data: 9237 1451 147 Winters et al. [11], Murzin et al. [12], Tsang and 9237 1451 148 Brugger [13], Smith [14], with level scheme from 9237 1451 149 Table of Isotopes [15] 9237 1451 150 Level No. Energy(MeV) Spin-parity 9237 1451 151 G.S. 0.0 0+ 9237 1451 152 1 0.0448 2+ 9237 1451 153 2 0.1484 4+ 9237 1451 154 3 0.3072 6+ 9237 1451 155 FISSION 9237 1451 156 Extrapolated downward from JEF-1. 9237 1451 157 CAPTURE CROSS-SECTION 9237 1451 158 Determined mainly by the post-1966 data compiled by 9237 1451 159 Poenitz [16], by the measurement of Kazakov [17] and 9237 1451 160 by the mean parameters (radiation width, level 9237 1451 161 density) extracted from resolved resonance 9237 1451 162 parameters [7]. 9237 1451 163 9237 1451 164 2200 M/S VALUES AND RESONANCE INTEGRALS GIVEN BY EVALUATION 9237 1451 165 9237 1451 166 2200m/s values Resonance Integral 9237 1451 167 (barns) (barns) 9237 1451 168 Total 12.32 597.4 9237 1451 169 Elastic 9.60 319.4 9237 1451 170 Fission 6.46E-05 1.7E-03 9237 1451 171 Capture 2.72 278.0 9237 1451 172 9237 1451 173 ---------------------------------------------------------------- 9237 1451 174 REFERENCES FOR MF=2 9237 1451 175 9237 1451 176 1. M.C. Moxon and M.G. Sowerby, Pr0c 1988 Int.Reac.Phys.Conf. 9237 1451 177 Jackson Hole (1988) Vol.1, p,281 9237 1451 178 2. R.L. Macklin et al, Conf. 1988 Mito, Japan, (1988) p.71 9237 1451 179 3. S.F. Mughabghab, Neutron Cross Sections, Vol.1, Part B, 9237 1451 180 Z=61-100 (Academic Press, 1984) 9237 1451 181 4. P. d'Hondt et al, Ann.Nucl.EN., 11, 485 (1984) 9237 1451 182 5. W.P. Poenitz, report IAEA-TECDOC-335 (1985) p.426 and 9237 1451 183 private communication (1987) 9237 1451 184 6. A.D. Carlson et al, Proc.Conf. 1985 Santa Fe, Vol.2, p.1429 9237 1451 185 (1985) 9237 1451 186 7. F.H. Froehner, PROC 1988 INT.REAC.PHYS. CONF., Jackson Hole 9237 1451 187 (1988) Vol.III, p 171 9237 1451 188 8. I. Tsubone et al, Nucl.Sci.Eng. 88, 579 (1984) 9237 1451 189 9. V.N. Kononov and E.D. Poletaev, 1973 KIEV CONF 2, 199 (1984)9237 1451 190 10. A.B. Smith et al, private communication (1988) 9237 1451 191 11. R.R. Winters et al, Nucl.Sci.Eng. 78, 147 (1981) 9237 1451 192 12. A.W. Murzin et al, Yad.Konst. 4, 30 (1986) 9237 1451 193 13. F.Y. Tsang and R.M. Brugger, Nucl.Sci.Eng. 65, 70 (1978) 9237 1451 194 14. A.B. Smith, report INDC/P(84)-28, unpublished (1984) 9237 1451 195 15. C.M. Lederer, Ed., Table of Isotopes, 7th edition, (1978) 9237 1451 196 16. W.P. Poenitz et al, Argonne National Lab. report ANL-83-4 9237 1451 197 suppl. (1983) 9237 1451 198 17. L.E. Kazakov et al, Yad.Konst. 3, 37 (1986) 9237 1451 199 18. F.C. Difilippo et al, Phys.Rev.C 21, 1400 (1980) 9237 1451 200 19. C.A. Uttley et al., Nucl.Data for Reactors, Proc.Int.Conf. 9237 1451 201 Paris 1966 (IAEA, 1966) Vol.I, p.165 9237 1451 202 20. T.Y. Byoun et al., PROC.MEET. REACTOR PHYS. AND SHIELD CALC.,9237 1451 203 Kiamesha Lake, NY, 1972 (U.S.Nucl.Reg.Comm., 1972) Vol.1, 9237 1451 204 p.1115 9237 1451 205 9237 1451 206 MF=3 SMOOTH CROSS SECTIONS 9237 1451 207 MT=102 Radiative Capture Cross Section. 10 to 149 keV, 9237 1451 208 evaluation by F.H. Froehner (KFK) which is based on FITACS 9237 1451 209 fit to the experimental data. 0.15 to 20 MeV, taken 9237 1451 210 directly from simultaneous standards analysis [Ca85], 9237 1451 211 [Po89]. 9237 1451 212 9237 1451 213 ---------------------------------------------------------------- 9237 1451 214 REFERENCES 9237 1451 215 9237 1451 216 [Al61] K. Allen et al., J.Nuc.En. 14, 100 (1961) 9237 1451 217 [An58] G. Antropov et al., At.En. 5, 456 (1958) 9237 1451 218 [Ar88] E.D. Arthur, Los Alamos National Lab. report LA-UR-88-382 9237 1451 219 (1988) 9237 1451 220 [Ba66] D. Barr, Los Aalamos National Lab., private communication 9237 1451 221 to R.Howerton (1966) 9237 1451 222 [Ba65] R. Batchelor et al., Nuc.Phys. 65, 236 (1965) 9237 1451 223 [Br58] A. Bratenahl et al., Phys.Rev. 110, 927 (1958) 9237 1451 224 [Ca73] J. Cabe et al., CEN Saclay report CEA-R-4524 (1973) 9237 1451 225 [Ca85] A. Carlson et al., Nucl. Data for Basic & Applied Science,9237 1451 226 Proc.Conf., Santa Fe, NM 1985 [] p.1429. 9237 1451 227 [Di80] F.C. Difillippo et al., Phys.Rev.C 21, 1400 (1980) 9237 1451 228 [Du70] C.L. Dunford, Atomics Int. report AI-AEC-12931 (1970) 9237 1451 229 [Fo71] D. Foster and D. Glasgow, Phys.Rev.C 3, 576 (1971) 9237 1451 230 [Fr80a] J. Frehaut et al., Nucl.Sci.Eng. 74, 29 (1980) 9237 1451 231 [Fr80b] J. Frehaut et al., Brookhaven National Lab. report 9237 1451 232 BNL-NCS-512457 (1980) p.399 9237 1451 233 [Fr86] J. Frehaut, NEANDC(E) 238/L (1986) 9237 1451 234 [Ha73] S. Hayes et al., Nuc.Sci.Eng. 50, 243 (1973) 9237 1451 235 [He80] D. Hetrick and C.Y. Fu, Oak Ridge National Lab. report 9237 1451 236 ORNL/TM-7341 (1980) 9237 1451 237 [Ka78] R. Kaiser and S. Carpenter (ANL-West) private 9237 1451 238 communication (1978). 9237 1451 239 [Ko80] N. Kornilov et al., report ZFK-410 (1980) p.68 9237 1451 240 [La73] J. Landrum et al., Phys.Rev.C 8, 1938 (1973) 9237 1451 241 [Li79] P. Lisowski et al., (LANL WNR measurement) private 9237 1451 242 communication (1979) 9237 1451 243 [Ma69] D. Mather, report AWRE-O-47 (1969) 9237 1451 244 [Ma72] D. Mather et al., report AWRE-O-72 (1972). 9237 1451 245 [Pe60] J. Peterson et al., Phys.Rev. 120, 521 (1960) 9237 1451 246 [Pe61] J. Perkin, J.Nuc.En. 14, 69 (1961) 9237 1451 247 [Ph56] J. Phillips, report AERE-NP/R-2033 (1956) 9237 1451 248 [Po81] W. Poenitz et al., Nuc.Sci.Eng. 78, 333 (1981) 9237 1451 249 [Po83] W.Poenitz et al., Argonne National Lab. report 9237 1451 250 ANL-NDM-80 (1983) 9237 1451 251 [Po89] W. Poenitz, ANL-West, private communication (1989) 9237 1451 252 [Ra70] J. Raynal, report IAEA SMR-9/8 (1970) 9237 1451 253 [Ro57] L. Rosen et al., Los Alamos National Lab. report LA-2111 9237 1451 254 (1958) 9237 1451 255 [Ry80] T. Ryves, J.Phys.G 6, 771 (1980) 9237 1451 256 [Sc74] R. Schwartz et al., Nuc.Sci.Eng. 54, 322 (1974) 9237 1451 257 [Sh78] R. Shamu et al., private communication (1978) 9237 1451 258 [Sl77] R. Slovacek et al., Nuc.Sci.Eng. 62, 455 (1977) 9237 1451 259 [Sm82] A. Smith et al., Argonne National Lab. report ANL/NDM-74 9237 1451 260 (1982) 9237 1451 261 [Ut66] C. Uttley et al., Paris Conf. (1966) vol.1, p.165 9237 1451 262 [Ve78] L. Veeser and E. Arthur, Harwell Nuc.Data Conf. (1978) 9237 1451 263 p 1054 9237 1451 264 [Wh62] P. White et al., J.Nuc.En.A/B 16, 261 (1962) 9237 1451 265 [Wh71] J. Whalen et al., Argonne National Lab. report ANL-7710 9237 1451 266 (1971) p.9 9237 1451 267 [Yo77] P.G. Young and E.D. Arthur, Los Alamos National Lab. 9237 1451 268 report LA-6947 (1977) 9237 1451 269 [Yo78] Chou You-Pu, report HSJ-77091 (1978) 9237 1451 270 9237 1451 271 **************************************************************** 9237 1451 272 Modifications applied by the IRDF evaluators 9237 1451 273 H.Vonach and S.Tagesen, IRK Vienna 9237 1451 274 MT=102 Capture Cross Section 9237 1451 275 Uncertainties from simultaneous standards analysis 9237 1451 276 used up to En=2.3+6 ev 9237 1451 277 Uncertainties for En=2.3+6 to 2.0+7 ev estimated from 9237 1451 278 existing experimental data (1) 9237 1451 279 REFERENCES: 9237 1451 280 1.Neutron Cross Sections, vol2, V.McLane, C.L.Dunford, 9237 1451 281 P.F.Rose, Academic Press(1988) 9237 1451 282 9237 1451 283 **************************************************************** 9237 1451 284 9237 1451 285 ENDF/B-V MAT 1393 Evaluation, November 1981 (ANL, LLNL) 9237 1451 286 9237 1451 287 Description of data carried over from ENDF/B-V 9237 1451 288 9237 1451 289 Principal Evaluators: 9237 1451 290 E.M.Pennington (ANL), A.B.Smith (ANL), W.P.Poenitz (ANL) 9237 1451 291 R.Howerton (LLL) - gamma production files 9237 1451 292 Contributing evaluators: 9237 1451 293 R.Kaiser and S.Carpenter (ANL-west) - nu-bar delayed and 9237 1451 294 delayed neutron spectra 9237 1451 295 9237 1451 296 MF = 3 smooth cross sections 9237 1451 297 Cross sections above 45 kev were evaluated by A.Smith. 9237 1451 298 more details are provided under each reaction type below. Most 9237 1451 299 of the evaluation above 45 kev is described in ANL/NDM-32. 9237 1451 300 ---------------------------------------------------------------- 9237 1451 301 REFERENCES 9237 1451 302 1. R.E. Kaiser and S.G. Carpenter (ANL-west) private comm. 9237 1451 303 (March 1978) 9237 1451 304 2. S.A. Cox, Argonne National Lab. report ANL/NDM-5 (1974) 9237 1451 305 3. C. Besant et al, Sem. Fast Pulsed Reactors CONF-760111 (1976)9237 1451 306 4. F. Manero and V. Konshin, At.En.Review 10, #4 (1972) 9237 1451 307 5. M. Soleilhac, revised data received from NNCSC 9237 1451 308 6. B. Nurpeisov et al., Sov.At.En. translation 807 (Mar 1976) 9237 1451 309 7. G. deSaussure et al., Prog.Nucl.En. 3, 87 (1979) 9237 1451 310 8. D. Olsen et al., Nuc.Sci.Eng. 62, 479 (1977) 9237 1451 311 9. P. Guenther, D. Havel, A. Smith Argonne National Lab. report 9237 1451 312 ANL-NDM-22 (1976) 9237 1451 313 10. E. Pennington, W. Poenitz, A. Smith, Trans.Am.Nucl.Soc. 26, 9237 1451 314 591 (1977) 9237 1451 315 11. W.P. Poenitz Argonne National Lab. report ANL/NDM-45 (1979) 9237 1451 316 12. W.P. Poenitz and A.B. Smith, ed. Argonne National Lab. report9237 1451 317 ANL-76-90 (1976) 9237 1451 318 13. R. Slovacek et al., Nucl.Sci.Eng. 62, 455 (1977) 9237 1451 319 14. M. Caner, M. Segev, W. Yiftah, Nucl.Sci.Eng. 59, 395 (1976) 9237 1451 320 15. E. Kujawski and L. Stewart, Trans.Am.Nucl.Soc. 24, 453 (1976)9237 1451 321 16. L.W. Weston private comm. to B.A. Magurno Nov.12, 1982 9237 1451 322 17. R. Sher and ?. Beck report EPRI NP-1771/81 and rev. 1/83, 9237 1451 323 and private communication to Magurno, February 1983 9237 1451 324 9237 1451 325 **************************************************************** 9237 1451 326 92-U -238 KYU,JAERI+ EVAL-MAR94 Y.KANDA ET AL. 9237 1451 327 DIST-JUL98 9237 1451 328 ----JENDL/D-99 MATERIAL 9237 9237 1451 329 -----INCIDENT NEUTRON DATA 9237 1451 330 ------ENDF-6 FORMAT 9237 1451 331 HISTORY 9237 1451 332 87-01 SIMULTANEOUS EVALUATION FOR FISSION AND CAPTURE CROSS 9237 1451 333 SECTIONS WAS COMPLETED IN THE ENERGY RANGE ABOVE 50 KEV. 9237 1451 334 93-03 FISSION SPECTRA CALCULATED BY T.OHSAWA(KINKI UNIV.) 9237 1451 335 COMPILED BY T.NAKAGAWA (NDC/JAERI) 9237 1451 336 98-04 COMPILED TO JENDL DOSIMETRY FILE 99. 9237 1451 337 9237 1451 338 ===== POINT-WISE DATA FILE ===== 9237 1451 339 9237 1451 340 9237 1451 341 U -238 FISSION 9237 1451 342 9237 1451 343 MF=1 GENERAL INFORMATION 9237 1451 344 MT=451 DESCRIPTIVE DATA AND DIRECTORY RECORDS 9237 1451 345 9237 1451 346 9237 1451 347 2200-M/S CROSS SECTIONS AND CALCULATED RESONANCE INTEGRALS. 9237 1451 348 2200 M/S(B) RES. INTEG.(B) 9237 1451 349 FISSION 11.8E-6 1.72 9237 1451 350 9237 1451 351 ***************** PROGRAM LINEAR (VERSION 87-1) *****************9237 1451 352 DATA LINEARIZED TO WITHIN AN ACCURACY OF 0.800 PER-CENT 9237 1451 353 ***************** PROGRAM RECENT (VERSION 89-1) *****************9237 1451 354 RESONANCE CONTRIBUTION RECONSTRUCTED TO WITHIN 1.000 PER-CENT 9237 1451 355 COMBINED DATA NOT THINNED (ALL RESONANCE + BACKGROUND DATA KEPT) 9237 1451 356 ***************** PROGRAM LINEAR (VERSION 87-1) *****************9237 1451 357 DATA LINEARIZED TO WITHIN AN ACCURACY OF 0.800 PER-CENT 9237 1451 358 9237 1451 359 MF=3 NEUTRON CROSS SECTIONS 9237 1451 360 BELOW 10 KEV, NO BACKGROUND CROSS SECTIONS WERE GIVEN. 9237 1451 361 ABOVE 10 KEV, CROSS SECTIONS WERE EVALUATED AS FOLLOWS, AND 9237 1451 362 THEY WERE REPRESENTED WITH THE UNRESOLVED RESONANCE 9237 1451 363 PARAMETERS EXCEPT THE FISSION CROSS SECTION. 9237 1451 364 9237 1451 365 MT=18 FISSION 9237 1451 366 BELOW 100 KEV : TAKEN FROM EXPERIMENTAL DATA /12/. 9237 1451 367 100 - 600 KEV : EVALUATED ON THE BASIS OF THE DATA OF 9237 1451 368 DIFFILIPPO ET AL. /13/, BEHRENS AND CARLSON /14/, 9237 1451 369 NORDBORG ET AL. /15/ AND MEADOWS /16,17/. 9237 1451 370 ABOVE 600 KEV : RESULTS OF SIMULTANEOUS EVALUATION /18/ MADE 9237 1451 371 BY CONSIDERING THE EXPERIMENTAL DATA OF REFS./14,15, 9237 1451 372 16,17,19,20,21,22,23,24,25,26,27,28/. 9237 1451 373 9237 1451 374 MF=33 9237 1451 375 MT=18 Taken from JENDL/D-99 9237 1451 376 9237 1451 377 9237 1451 378 REFERENCES 9237 1451 379 12) DIFILLIPO F.C. ET AL.: NUCL. SCI. ENG., 63, 153 (1977). 9237 1451 380 13) DIFILLIPO F.C. ET AL.: NUCL. SCI. ENG., 68, 43 (1978). 9237 1451 381 14) BEHRENS J.W. AND CARLSON G.W.: NUCL. SCI. ENG., 63, 250 9237 1451 382 (1977). 9237 1451 383 15) NORDBORG C. ET AL: ANL-76-90, 128 (1976). 9237 1451 384 16) MEADOWS J.W.: NUCL. SCI. ENG., 58, 255 (1975). 9237 1451 385 17) MEADOWS J.W.: NUCL. SCI. ENG., 49, 310 (1972). 9237 1451 386 18) KANDA Y. ET AL.: 1985 SANTA FE, 2, 1567 (1986). 9237 1451 387 19) CANCE M. AND GRENIER G.:NUCL. SCI. ENG., 68, 197 (1978). 9237 1451 388 20) BILAUD P. ET AL.:1958 GENEVA,16, 106, 5809 (1958). 9237 1451 389 21) ADAMOV V.M. ET AL.: 1977 NBS, 313 (1977). 9237 1451 390 22) ARLT R. ET AL.: KE, 24, 48, 8102 (1981). 9237 1451 391 23) CIERJACKS S. ET AL.:1976 ANL, 94 (1976). 9237 1451 392 24) GOVERDOVSKII A.A. ET AL.: 1983 KIEV, 2, 159 (1983). 9237 1451 393 25) ANDROSENKO S.D. ET AL.: 1983 KIEV, 2, 153 (1983). 9237 1451 394 26) FURSOV B.I. ET AL.: SOV. ATOM. ENERG., 43, 808 (1978). 9237 1451 395 27) POENITZ W.P. AND ARMANI R.J.: J. NUCL. ENEG., 26, 483 (1972). 9237 1451 396 28) POENITZ W.P.: NUCL. SCI. ENG., 57, 300 (1975). 9237 1451 397 9237 1451 398 ************************ C O N T E N T S *********************** 9237 1451 399 ***************** Program LINEAR (VERSION 2002-1) ***************9237 1451 400 For All Data Greater than 1.0000E-10 barns in Absolute Value 9237 1451 401 Data Linearized to Within an Accuracy of .100000000 per-cent 9237 1451 402 ***************** Program RECENT (VERSION 2002-1) ***************9237 1451 403 for All Data Greater than 1.0000E-10 barns in Absolute Value 9237 1451 404 Data Linearized to within an Accuracy of .100000000 per-cent 9237 1451 405 ***************** Program SIGMA1 (VERSION 2002-1) ***************9237 1451 406 Data Doppler Broadened to 300.000000 Kelvin 9237 1451 407 for All Data Greater than 1.0000E-10 barns in Absolute Value 9237 1451 408 Data Linearized to Within an Accuracy pf .100000000 per-cent 9237 1451 409 ***************** Program FIXUP (Version 2002-1) ****************9237 1451 410 Corrected ZA/AWR in All Sections-----------------------------Yes 9237 1451 411 Corrected Thresholds-----------------------------------------Yes 9237 1451 412 Extended Cross Sections to 20 MeV----------------------------No 9237 1451 413 Allow Cross Section Deletion---------------------------------No 9237 1451 414 Allow Cross Section Reconstruction---------------------------No 9237 1451 415 Make All Cross Sections Non-Negative-------------------------Yes 9237 1451 416 Delete Energies Not in Ascending Order-----------------------Yes 9237 1451 417 Deleted Duplicate Points-------------------------------------Yes 9237 1451 418 Check for Ascending MAT/MF/MT Order--------------------------Yes 9237 1451 419 Check for Legal MF/MT Numbers--------------------------------Yes 9237 1451 420 Allow Creation of Missing Sections---------------------------No 9237 1451 421 Allow Insertion of Energy Points-----------------------------No 9237 1451 422 Create Uniform Energy Grid-----------------------------------No 9237 1451 423 Delete Section if Cross Section =0 at All Energies-----------Yes 9237 1451 424 ***************** Program GROUPIE (VERSION 2002-1) **************9237 1451 425 Unshielded Group Averages Using 640 Groups 9237 1451 426 Weighting Spectrum: Flat (Constant) Spectrum 9237 1451 427 1 451 433 49237 1451 428 2 151 4 49237 1451 429 3 18 217 19237 1451 430 3 102 217 39237 1451 431 33 18 11 09237 1451 432 33 102 335 09237 1451 433 9237 1 099999 9237 0 0 0 9.22380E+4 2.36005E+2 0 0 1 09237 2151 1 9.22380E+4 1.00000E+0 0 0 1 09237 2151 2 1.00000E-5 1.00000E+4 0 0 0 09237 2151 3 0.0 9.42848E-1 0 0 0 09237 2151 4 9237 2 099999 9237 0 0 0 9.22380E+4 2.36005E+2 0 0 0 09237 3 18 1 1.89629E+8 1.89629E+8 0 0 1 6419237 3 18 2 641 1 9237 3 18 3 .000100000 .000184668 .000105000 .000180362 .000110000 .0001762729237 3 18 4 .000115000 .000172502 .000120000 .000168096 .000127500 .0001632069237 3 18 5 .000135000 .000158740 .000142500 .000154635 .000150000 .0001501939237 3 18 6 .000160000 .000145564 .000170000 .000141353 .000180000 .0001374889237 3 18 7 .000190000 .000133892 .000200000 .000130609 .000210000 .0001275209237 3 18 8 .000220000 .000124650 .000230000 .000121984 .000240000 .0001188559237 3 18 9 .000255000 .000115416 .000270000 .000112762 .000280000 .0001098089237 3 18 10 .000300000 .000106207 .000320000 .000102945 .000340000 9.99571E-59237 3 18 11 .000360000 9.72059E-5 .000380000 9.46943E-5 .000400000 9.20757E-59237 3 18 12 .000425000 8.93960E-5 .000450000 8.69475E-5 .000475000 8.47028E-59237 3 18 13 .000500000 8.25920E-5 .000525000 8.06627E-5 .000550000 7.88401E-59237 3 18 14 .000575000 7.71450E-5 .000600000 7.54077E-5 .000630000 7.36205E-59237 3 18 15 .000660000 7.19851E-5 .000690000 7.04164E-5 .000720000 6.87471E-59237 3 18 16 .000760000 6.69572E-5 .000800000 6.52954E-5 .000840000 6.37724E-59237 3 18 17 .000880000 6.23249E-5 .000920000 6.09980E-5 .000960000 5.97271E-59237 3 18 18 .001000000 5.84097E-5 .001050000 5.70345E-5 .001100000 5.57454E-59237 3 18 19 .001150000 5.45576E-5 .001200000 5.31564E-5 .001275000 5.16164E-59237 3 18 20 .001350000 5.02084E-5 .001425000 4.88945E-5 .001500000 4.74987E-59237 3 18 21 .001600000 4.60394E-5 .001700000 4.47074E-5 .001800000 4.34754E-59237 3 18 22 .001900000 4.23502E-5 .002000000 4.13069E-5 .002100000 4.03278E-59237 3 18 23 .002200000 3.94312E-5 .002300000 3.85737E-5 .002400000 3.75953E-59237 3 18 24 .002550000 3.65078E-5 .002700000 3.56570E-5 .002800000 3.47323E-59237 3 18 25 .003000000 3.35937E-5 .003200000 3.25576E-5 .003400000 3.16104E-59237 3 18 26 .003600000 3.07471E-5 .003800000 2.99506E-5 .004000000 2.91208E-59237 3 18 27 .004250000 2.82791E-5 .004500000 2.75109E-5 .004750000 2.67928E-59237 3 18 28 .005000000 2.61334E-5 .005250000 2.55216E-5 .005500000 2.49428E-59237 3 18 29 .005750000 2.44116E-5 .006000000 2.38534E-5 .006300000 2.32950E-59237 3 18 30 .006600000 2.27684E-5 .006900000 2.22829E-5 .007200000 2.17535E-59237 3 18 31 .007600000 2.11862E-5 .008000000 2.06673E-5 .008400000 2.01803E-59237 3 18 32 .008800000 1.97270E-5 .009200000 1.93056E-5 .009600000 1.89046E-59237 3 18 33 .010000000 1.84894E-5 .010500000 1.80505E-5 .011000000 1.76491E-59237 3 18 34 .011500000 1.72663E-5 .012000000 1.68277E-5 .012750000 1.63424E-59237 3 18 35 .013500000 1.58931E-5 .014250000 1.54812E-5 .015000000 1.50421E-59237 3 18 36 .016000001 1.45812E-5 .017000001 1.41583E-5 .017999999 1.37719E-59237 3 18 37 .018999999 1.34166E-5 .020000000 1.30832E-5 .021000000 1.27781E-59237 3 18 38 .022000000 1.24907E-5 .023000000 1.22230E-5 .024000000 1.19084E-59237 3 18 39 .025500000 1.15692E-5 .027000001 1.13006E-5 .028000001 1.10089E-59237 3 18 40 .029999999 1.06500E-5 .032000002 1.03234E-5 .034000002 1.00257E-59237 3 18 41 .035999998 9.75425E-6 .037999999 9.50355E-6 .039999999 9.24161E-69237 3 18 42 .042500000 8.97539E-6 .045000002 8.73211E-6 .047499999 8.50551E-69237 3 18 43 .050000001 8.29718E-6 .052499998 8.10449E-6 .055000000 7.92212E-69237 3 18 44 .057500001 7.75531E-6 .059999999 7.58005E-6 .063000001 7.40522E-69237 3 18 45 .066000000 7.24071E-6 .068999998 7.08773E-6 .071999997 6.92131E-69237 3 18 46 .075999998 6.74264E-6 .079999998 6.57928E-6 .083999999 6.42576E-69237 3 18 47 .088000000 6.28307E-6 .092000000 6.15099E-6 .096000001 6.02488E-69237 3 18 48 .100000001 5.89475E-6 .104999997 5.75707E-6 .109999999 5.63137E-69237 3 18 49 .115000002 5.51135E-6 .119999997 5.37402E-6 .127499998 5.22263E-69237 3 18 50 .135000005 5.08266E-6 .142499998 4.95465E-6 .150000006 4.81743E-69237 3 18 51 .159999996 4.67296E-6 .170000002 4.54092E-6 .180000007 4.42044E-69237 3 18 52 .189999998 4.30995E-6 .200000003 4.20631E-6 .209999993 4.11165E-69237 3 18 53 .219999999 4.02241E-6 .230000004 3.93941E-6 .239999995 3.84384E-69237 3 18 54 .254999995 3.73714E-6 .270000011 3.65553E-6 .280000001 3.56497E-69237 3 18 55 .300000012 3.45451E-6 .319999993 3.35535E-6 .340000004 3.26599E-69237 3 18 56 .360000014 3.18067E-6 .379999995 3.10393E-6 .400000006 3.02593E-69237 3 18 57 .425000012 2.94428E-6 .449999988 2.87043E-6 .474999994 2.80166E-69237 3 18 58 .500000000 2.73988E-6 .524999976 2.68321E-6 .550000012 2.62772E-69237 3 18 59 .574999988 2.57815E-6 .600000024 2.52624E-6 .629999995 2.47453E-69237 3 18 60 .660000026 2.42564E-6 .689999998 2.38131E-6 .720000029 2.33401E-69237 3 18 61 .759999990 2.28178E-6 .800000012 2.23587E-6 .839999974 2.19123E-69237 3 18 62 .879999995 2.15172E-6 .920000017 2.11434E-6 .959999979 2.08108E-69237 3 18 63 1.00000000 2.04424E-6 1.04999995 2.00718E-6 1.10000002 1.97269E-69237 3 18 64 1.14999998 1.94246E-6 1.20000005 1.90602E-6 1.27499998 1.86651E-69237 3 18 65 1.35000002 1.83288E-6 1.42499995 1.80086E-6 1.50000000 1.76880E-69237 3 18 66 1.60000002 1.73647E-6 1.70000005 1.70830E-6 1.79999995 1.68472E-69237 3 18 67 1.89999998 1.66301E-6 2.00000000 1.64531E-6 2.09999990 1.63059E-69237 3 18 68 2.20000005 1.61794E-6 2.29999995 1.60871E-6 2.40000010 1.59963E-69237 3 18 69 2.54999995 1.59324E-6 2.70000005 1.59020E-6 2.79999995 1.59329E-69237 3 18 70 3.00000000 1.60347E-6 3.20000005 1.62120E-6 3.40000010 1.64925E-69237 3 18 71 3.59999990 1.68712E-6 3.79999995 1.73754E-6 4.00000000 1.81176E-69237 3 18 72 4.25000000 1.92264E-6 4.50000000 2.07439E-6 4.75000000 2.28731E-69237 3 18 73 5.00000000 2.59737E-6 5.25000000 3.07418E-6 5.50000000 3.87484E-69237 3 18 74 5.75000000 5.42717E-6 6.00000000 1.02457E-5 6.30000019 .0001078319237 3 18 75 6.59999990 .001115810 6.90000010 8.67297E-6 7.19999981 1.30114E-69237 3 18 76 7.59999990 3.80596E-7 8.00000000 2.50553E-7 8.39999962 2.58209E-79237 3 18 77 8.80000019 2.97648E-7 9.19999981 3.43688E-7 9.60000038 3.90050E-79237 3 18 78 10.0000000 4.40501E-7 10.5000000 4.95398E-7 11.0000000 5.49559E-79237 3 18 79 11.5000000 6.04128E-7 12.0000000 6.76534E-7 12.7500000 7.73992E-79237 3 18 80 13.5000000 8.90463E-7 14.2500000 1.03734E-6 15.0000000 1.27641E-69237 3 18 81 16.0000000 1.70901E-6 17.0000000 2.53130E-6 18.0000000 4.53423E-69237 3 18 82 19.0000000 1.29072E-5 20.0000000 .003044283 21.0000000 .0002335999237 3 18 83 22.0000000 6.21566E-6 23.0000000 2.11257E-6 24.0000000 1.00282E-69237 3 18 84 25.5000000 6.22895E-7 27.0000000 5.34668E-7 28.0000000 5.45691E-79237 3 18 85 30.0000000 6.90426E-7 32.0000000 1.16027E-6 34.0000000 4.80306E-69237 3 18 86 36.0000000 .000321611 38.0000000 6.81239E-7 40.0000000 8.15589E-89237 3 18 87 42.5000000 7.06497E-8 45.0000000 1.10208E-7 47.5000000 1.58009E-79237 3 18 88 50.0000000 2.13391E-7 52.5000000 2.84426E-7 55.0000000 3.90089E-79237 3 18 89 57.5000000 5.84208E-7 60.0000000 1.21992E-6 63.0000000 .0002314589237 3 18 90 66.0000000 .000328269 69.0000000 4.80554E-7 72.0000000 2.11524E-79237 3 18 91 76.0000000 5.97862E-7 80.0000000 5.74436E-5 84.0000000 5.16331E-89237 3 18 92 88.0000000 1.16773E-7 92.0000000 2.17638E-7 96.0000000 5.95206E-79237 3 18 93 100.000000 7.71072E-5 105.000000 2.38574E-7 110.000000 8.34158E-89237 3 18 94 115.000000 8.57761E-8 120.000000 1.03492E-7 127.500000 1.24924E-79237 3 18 95 135.000000 1.47168E-7 142.500000 1.71946E-7 150.000000 2.09340E-79237 3 18 96 160.000000 2.77760E-7 170.000000 4.55506E-7 180.000000 5.95252E-59237 3 18 97 190.000000 9.40205E-6 200.000000 .000111774 210.000000 4.93613E-79237 3 18 98 220.000000 6.74250E-8 230.000000 3.22793E-5 240.000000 6.33722E-89237 3 18 99 255.000000 8.06979E-8 270.000000 1.08109E-7 280.000000 1.43942E-79237 3 18 100 300.000000 2.15402E-7 320.000000 5.05216E-7 340.000000 .0001076249237 3 18 101 360.000000 3.66765E-6 380.000000 8.82725E-8 400.000000 1.65882E-79237 3 18 102 425.000000 2.95672E-7 450.000000 9.49531E-5 475.000000 1.21865E-59237 3 18 103 500.000000 5.03088E-5 525.000000 7.41218E-5 550.000000 3.70682E-79237 3 18 104 575.000000 .000220720 600.000000 1.84857E-5 630.000000 3.58844E-79237 3 18 105 660.000000 1.60780E-6 690.000000 .003119950 720.000000 .0156989949237 3 18 106 760.000000 1.40855E-7 800.000000 2.24445E-5 840.000000 .0003646309237 3 18 107 880.000000 1.61614E-7 920.000000 1.63494E-8 960.000000 4.28418E-99237 3 18 108 1000.00000 4.45165E-8 1050.00000 2.45123E-7 1100.00000 .0001542799237 3 18 109 1150.00000 .000867571 1200.00000 .004540664 1275.00000 3.63512E-79237 3 18 110 1350.00000 1.33179E-7 1425.00000 6.97851E-8 1500.00000 3.92568E-89237 3 18 111 1600.00000 2.28333E-8 1700.00000 1.42444E-8 1800.00000 9.50553E-99237 3 18 112 1900.00000 6.34242E-9 2000.00000 4.49933E-9 2100.00000 3.19030E-99237 3 18 113 2200.00000 2.34720E-9 2300.00000 1.78264E-9 2400.00000 1.31466E-99237 3 18 114 2550.00000 1.00839E-9 2700.00000 9.1651E-10 2800.00000 9.0786E-109237 3 18 115 3000.00000 1.03038E-9 3200.00000 1.26485E-9 3400.00000 1.57850E-99237 3 18 116 3600.00000 1.99590E-9 3800.00000 2.50887E-9 4000.00000 3.24264E-99237 3 18 117 4250.00000 4.25527E-9 4500.00000 5.60104E-9 4750.00000 7.59153E-99237 3 18 118 5000.00000 1.09465E-8 5250.00000 1.87454E-8 5500.00000 9.98495E-59237 3 18 119 5750.00000 2.07209E-8 6000.00000 1.34150E-8 6300.00000 3.06706E-89237 3 18 120 6600.00000 7.54352E-8 6900.00000 .000249709 7200.00000 .0004390949237 3 18 121 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1.756601-8 2.307246-8 3.049692-8 6.589630-8923733102 286 4.306795-8 3.891639-8 9.316174-8 6.233225-8 5.014352-8 1.132461-7923733102 287 7.828227-8 7.217931-8 7.466687-8 8.420934-8 8.280880-8 7.986202-8923733102 288 7.099759-8 6.653936-8 3.729051-8 6.229219-8 1.756601-8 1.756601-8923733102 289 2.307246-8 3.049692-8 6.589630-8 4.306795-8 3.891639-8 9.316174-8923733102 290 6.233225-8 5.014352-8 1.132461-7 7.828227-8 7.217931-8 7.466687-8923733102 291 8.420934-8 8.280880-8 7.986202-8 7.099759-8 6.653936-8 3.729051-8923733102 292 6.229219-8 4.462786-7 4.462786-7 2.837186-6 2.806743-6 3.223859-6923733102 293 2.752009-6 3.010122-6 2.961952-6 2.971455-6 3.034000-6 2.467666-6923733102 294 2.328588-6 2.168345-6 2.148547-6 2.101845-6 1.956477-6 1.868093-6923733102 295 1.606852-6 1.416914-6 7.693215-7 1.329355-6 4.845855-7 4.845855-7923733102 296 3.132689-6 3.256939-6 3.913624-6 3.392811-6 3.695923-6 3.566611-6923733102 297 3.470179-6 3.539297-6 2.892230-6 2.746992-6 2.550140-6 2.519340-6923733102 298 2.457250-6 2.281159-6 2.177644-6 1.876005-6 1.653521-6 8.985493-7923733102 299 1.551922-6 5.522040-7 5.522040-7 3.438471-6 3.867718-6 6.089964-6923733102 300 4.935918-6 5.382186-6 4.995669-6 4.606240-6 4.582258-6 3.812452-6923733102 301 3.710496-6 3.431793-6 3.369575-6 3.273834-6 3.032319-6 2.894236-6923733102 302 2.499881-6 2.202559-6 1.199479-6 2.068463-6 5.048252-7 5.048252-7923733102 303 3.168782-6 3.887495-6 5.527578-6 7.574905-6 8.200554-6 7.254595-6923733102 304 5.857552-6 5.588665-6 4.658222-6 4.595782-6 4.214876-6 4.082617-6923733102 305 3.923727-6 3.600930-6 3.433465-6 2.983538-6 2.618851-6 1.420304-6923733102 306 2.460591-6 4.740287-7 4.740287-7 3.176372-6 3.970602-6 5.519170-6923733102 307 7.847616-6 1.357481-5 9.459232-6 6.909114-6 6.411306-6 5.445994-6923733102 308 5.421315-6 4.991993-6 4.829055-6 4.591193-6 4.185679-6 3.990721-6923733102 309 3.479556-6 3.045590-6 1.645466-6 2.863388-6 4.904450-7 4.904450-7923733102 310 3.224962-6 3.648894-6 4.494154-6 6.372683-6 8.250372-6 1.645021-5923733102 311 8.292258-6 6.662842-6 5.744732-6 5.594043-6 5.180347-6 5.028297-6923733102 312 4.732302-6 4.281294-6 4.081069-6 3.583898-6 3.108881-6 1.671810-6923733102 313 2.921791-6 5.179343-7 5.179343-7 3.228833-6 3.502100-6 4.376715-6923733102 314 5.429168-6 6.657624-6 8.720918-6 1.189946-5 7.362235-6 6.140660-6923733102 315 5.733672-6 5.281981-6 5.181792-6 4.906823-6 4.452004-6 4.239362-6923733102 316 3.730595-6 3.251817-6 1.751799-6 3.052687-6 4.618064-7 4.618064-7923733102 317 3.101498-6 3.554217-6 4.180992-6 4.990470-6 5.929547-6 6.708322-6923733102 318 7.279768-6 1.250699-5 7.214284-6 6.057901-6 5.665926-6 5.554190-6923733102 319 5.173895-6 4.593276-6 4.355488-6 3.817554-6 3.360511-6 1.809253-6923733102 320 3.170271-6 4.124357-7 4.124357-7 2.762236-6 3.088725-6 3.571202-6923733102 321 4.445007-6 5.195247-6 6.309300-6 6.113009-6 7.239937-6 1.216836-5923733102 322 6.836683-6 5.833379-6 5.606253-6 5.194231-6 4.520603-6 4.285438-6923733102 323 3.770336-6 3.302310-6 1.771589-6 3.122687-6 2.810233-7 2.810233-7923733102 324 2.155689-6 2.505228-6 2.936489-6 4.045053-6 4.836048-6 5.855992-6923733102 325 5.702776-6 5.664378-6 6.012916-6 1.167280-5 7.220186-6 6.264512-6923733102 326 5.695446-6 4.919647-6 4.676586-6 4.184389-6 3.636112-6 1.925731-6923733102 327 3.444378-6 1.203301-7 1.203301-7 1.262618-6 1.521675-6 1.824114-6923733102 328 2.555311-6 2.803257-6 3.856687-6 3.718516-6 4.195658-6 4.391636-6923733102 329 6.740464-6 1.247719-5 8.932404-6 6.456683-6 5.685773-6 5.465747-6923733102 330 4.982233-6 4.350095-6 2.231388-6 4.170753-6 6.444943-8 6.444943-8923733102 331 9.519869-7 1.166006-6 1.423474-6 2.034209-6 2.151504-6 3.242044-6923733102 332 3.021133-6 3.535045-6 3.502486-6 5.328298-6 6.687729-6 1.212462-5923733102 333 9.427675-6 6.431210-6 6.282109-6 5.700808-6 4.973386-6 2.534591-6923733102 334 4.757867-6 923733102 335 923733 099999 9237 0 0 0 0 0 0 0 9.32370E+4 2.35011E+2 0 1 34 109346 1451 1 0.0 1.00000E+0 0 0 0 69346 1451 2 1.00000E+0 2.00000E+7 0 0 10 20029346 1451 3 3.00000E+2 0.0 1 0 345 49346 1451 4 93-Np-237 FEI/LANL EVAL-May96 K.Zolotarev et al.,P.Young et al.9346 1451 5 DIST-Feb2004 9346 1451 6 ----IRDF-2002 MATERIAL 9346 9346 1451 7 -----INCIDENT NEUTRON DATA 9346 1451 8 ------ENDF-6 FORMAT 9346 1451 9 ***************************************************************** 9346 1451 10 93-Np-237 FEI EVAL-May96 K.Zolotarev et al. 9346 1451 11 DIST-Sep03 9346 1451 12 ----BROND-3 MATERIAL 9346 Revision 2, Oct. 2003 9346 1451 13 ***************************************************************** 9346 1451 14 ------Russian Reactor Dosimetry File RRDF-2002 9346 1451 15 ***************************************************************** 9346 1451 16 Authors of evaluation: K.I.Zolotarev, S.A.Badikov, A.I.Blokhin, 9346 1451 17 A.V.Ignatyuk 9346 1451 18 ***************************************************************** 9346 1451 19 9346 1451 20 ----- MF=2 MT=151 ----- 9346 1451 21 9346 1451 22 RESONANCE PARAMETERS 9346 1451 23 -------------------- 9346 1451 24 Resolved Resonance Region : 1.0E-5 - 600 eV 9346 1451 25 Resolved MLBW resonance parameters up to 600 eV were based on 9346 1451 26 on data from ref. [1]. 9346 1451 27 A statistical method of the resonance analysis showed that 9346 1451 28 the missing of the resonances becomes essential above the energy 9346 1451 29 of 150 eV. However only weak resonances are missed with the neut- 9346 1451 30 ron widths less than the average width by the factor 5-10. 9346 1451 31 The resonances with widths close to average or above it were 9346 1451 32 identified without the noticeable missing up to the energy of 9346 1451 33 600 eV. So as these resonances give dominant contribution into 9346 1451 34 the neutron cross sections the resolved resonances region was ex- 9346 1451 35 panded up to 600 eV and average contribution of the missed weak 9346 1451 36 resonances was taken into account by addition of relevant cross 9346 1451 37 section background in the file MF=3. (For fission cross sections 9346 1451 38 this background is equal to zero). 9346 1451 39 9346 1451 40 Unresolved Resonance Region : 600 eV - 6 keV 9346 1451 41 For the optical-statistical calculation of the cross sections 9346 1451 42 in the unresolved resonance region were used the following para- 9346 1451 43 meters: 9346 1451 44 Average resonance spacing - D0 = 0.57+-0.03 eV 9346 1451 45 Neutron strengh function - S0 = (0.97+-0.07)*104 9346 1451 46 Average radiation width - Gg = 40.0+-1.2 meV 9346 1451 47 This values were obtained as a result of the analysis resolwed 9346 1451 48 MLBW resonance parameters taking into account the missed resonan- 9346 1451 49 ces correction. 9346 1451 50 Considerable attention was paid to the analysis of fission 9346 1451 51 widths and elimination of contradictions in the description of 9346 1451 52 intermediate structure of averaged fission cross sections for the 9346 1451 53 neutron energies above 100 eV. The new experimental data [2-4] on 9346 1451 54 fission cross sections were included in the analysis. They were 9346 1451 55 obtained by LANL and JAERI physicists on their neutron spectrome- 9346 1451 56 ter [2,3] and by Dubna-Obninsk collaboration on the pulsed reac- 9346 1451 57 tor of JINR [4]. 9346 1451 58 The resulting set of resonance parameters reproduces well the 9346 1451 59 observed intermediate structure of fission cross sections (100 eV-9346 1451 60 6 keV). Corresponding group fission cross sections are 30-50 % 9346 1451 61 lower than JENDL-3.2 evaluation [5] and approximately twice 9346 1451 62 higher than ENDF/B-VI ones [6]. Corrected fission widths of the 9346 1451 63 neutron resonances below 100 eV also result in fission cross sec- 9346 1451 64 tions which are lower than JENDL-3.2 values [5]. 9346 1451 65 Calculated from the adopted resonance parameters 2200 M/S 9346 1451 66 cross sections and resonance integrals are agree rather well with 9346 1451 67 the recommended evaluated experimental data [1] : 9346 1451 68 9346 1451 69 Value This evaluation Experimental data [1] 9346 1451 70 Total 190.23 barn 9346 1451 71 Elastic 14.17 barn 9346 1451 72 Fission 0.0215 barn 0.0215+-0.0024 barn 9346 1451 73 Capture 176.04 barn 175.9+-2.9 barn 9346 1451 74 Fission res. int. 6.930 barn 6.9+-1.0 barn 9346 1451 75 Capture res. int. 642.30 barn 640+-50 barn 9346 1451 76 9346 1451 77 ----- MF=3 MT= 18 ----- 9346 1451 78 9346 1451 79 FISSION CROSS SECTIONS 9346 1451 80 ---------------------- 9346 1451 81 Np-237 fission cross sections from 1.0E-5 eV to 6 keV are re- 9346 1451 82 constructed from evaluated MLBW resolved and unresolved resonance 9346 1451 83 parameters. 9346 1451 84 For the evaluation of the Np-237(n,f) cross sections in the 9346 1451 85 neutron energy region of 6 keV - 20 MeV available experimental 9346 1451 86 data [7- 46] were analyzed. During this procedure experimental 9346 1451 87 data if it was possible were corrected to the new recommended 9346 1451 88 cross section data for monitor reactions used in the measurements 9346 1451 89 and to the new recommended decay data. 9346 1451 90 The top priority was given to absolute measurements where no 9346 1451 91 reference cross sections were used to determine the neutron flux 9346 1451 92 and to time-of-flight experiments with simultaneous registration 9346 1451 93 of the fission and monitoring reaction events. 9346 1451 94 In ten experiments the Np-237(n,f) cross sections were measu- 9346 1451 95 red relatively to the fission of U-235. The use as the standard 9346 1451 96 of the U-235 fission cross section from ENDF/B-VI [47] instead 9346 1451 97 of the old one (ENDF/B-V) results in decrease of the Np-237(n,f) 9346 1451 98 cross sections in average by 2 % in 0.1 - 2.0 MeV range and by 9346 1451 99 1.5 % in 2.0 - 3.0 MeV range. 9346 1451 100 In the work of Kupriyanov et al. [20] the Np-237(n,f) cross 9346 1451 101 sections were measured against the fission cross sections of 9346 1451 102 Pu-239. At the present time there are no recommended Pu-239 cross 9346 1451 103 sections as a standard data. The data from two libraries were 9346 1451 104 used to get absolute values: JENDL-3.2 [48] and ENDF/B-VI [49]. 9346 1451 105 Below 1.6 MeV the results of Ref. [20] disagree with the integral 9346 1451 106 experiments no matter which data are used for Pu-239(n,f) . The 9346 1451 107 analysis of Ref. [20] data demonstrated that the good agreement 9346 1451 108 with the rest of data may be obtained for the energy region of 9346 1451 109 1.6-7.0 MeV if Pu-239(n,f) reaction cross section data are taken 9346 1451 110 from JENDL-3.2 library. 9346 1451 111 The ratio of Np-237 to U-235 fission cross sections measured 9346 1451 112 by Behrens et al. [23] in 0.11 - 18.89 MeV energy range was multi-9346 1451 113 plied by factor of 1.051. This normalization factor was obtained 9346 1451 114 from the values of the functional / in 9346 1451 115 the 1 - 5 MeV energy range evaluated before. There is a lot of 9346 1451 116 experimental data which are in good agreement in this interval. 9346 1451 117 The ratio of the Np-237(n,f) evaluated averaged cross section for 9346 1451 118 the neutron spectrum of Cf-252 spontaneous fission, known from 9346 1451 119 many works, to the averaged data of Behrens et al. [23] is equal 9346 1451 120 to 1.055 that confirms our renormalization of these data. 9346 1451 121 The analysis of experimental data on the ratio of Np-237 and 9346 1451 122 U-235 fission cross sections indicates that the relative energy 9346 1451 123 trends of / measured by Terayama et al.9346 1451 124 [36] in the energy range of 4.19 - 6.99 MeV and by Goverdovskiy 9346 1451 125 et al. [32] in the 5.66 - 10.06 MeV energy range coincides with 9346 1451 126 the results of other authors. To make them agree in absolute 9346 1451 127 values Terayama et al. and Goverdovskiy et al. results were multi-9346 1451 128 plied by 0.96 and 1.079, respectively. The data on the Np-237 9346 1451 129 fission cross sections obtained by Terayama et al. on T(p,n)He-3 9346 1451 130 neutron source for the 0.70 - 2.99 MeV energy range are in good 9346 1451 131 agreement with the results of Refs. [38-40]. So they were correc- 9346 1451 132 ted only according to the new data on the monitoring reaction 9346 1451 133 U-235(n,f) [46]. 9346 1451 134 Experimental data of Meadows et al. [26] as well as that of 9346 1451 135 Kupriyanov et al. [20] are systematically too low below 1 MeV 9346 1451 136 that contradicts to the evaluated integral experiments 9346 1451 137 available. Above 1 MeV Meadows et al. [26] data well agree with 9346 1451 138 the results of other authors so they were included in the final 9346 1451 139 evaluation only above this energy. Data of Kupriyanov et al. [20] 9346 1451 140 were taken into account in the evaluation above 1.6 MeV. 9346 1451 141 Statistical analysis of the input experimental cross section 9346 1451 142 data for Np-237(n,f) reaction in 0.1-20 MeV neutron energy range 9346 1451 143 was carried out by means of PADE-2 code [50]. Rational function 9346 1451 144 was used as the model function [51]. 9346 1451 145 Averaged cross sections for U-235 thermal fission [52] and 9346 1451 146 Cf-252 spontaneous fission neutron spectra [53] calculated from 9346 1451 147 the evaluated Np-237(n,f) excitation function are the following: 9346 1451 148 9346 1451 149 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÂÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 9346 1451 150 TYPE OF SPECTRUM ³ ,mb (calc.) ³ , mb (measured) 9346 1451 151 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 9346 1451 152 U-235 neutron fission ³ 1356.2 ³ 1353.0 +- 24.0 [54] 9346 1451 153 ³ ³ 1350.0 +- 24.0 [55] 9346 1451 154 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÅÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 9346 1451 155 CF-252 spont. fission ³ 1359.9 ³ 1361.0 +- 21.6 [55] 9346 1451 156 ÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÁÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄÄ 9346 1451 157 9346 1451 158 Evaluated Np-237 fission cross section is in a good agreement 9346 1451 159 with the integral experiment data both for U-235 thermal fission 9346 1451 160 neutron spectrum and for Cf-252 spontaneous fission neutron spec- 9346 1451 161 trum. 9346 1451 162 The evaluated Np-237(n,f) cross section agrees within the ac- 9346 1451 163 curacy with integral experimental data for SIGNA-SIGMA field. 9346 1451 164 The Np-237 fission cross sections measured in CFRMF facility 9346 1451 165 are lower than calculated ones for any fission cross section eva- 9346 1451 166 luation. This apparently indicates the necessity of more detailed 9346 1451 167 and careful analysis of the CFRMF experimental data accuracy. 9346 1451 168 9346 1451 169 ----- MF=33 MT= 18----- 9346 1451 170 Uncertainties in the evaluated excitation function for the 9346 1451 171 Np-237(n,f) reaction are given in the two independent matrixes. 9346 1451 172 In the energy range 1.000E-05 eV - 0.1 MeV uncertainties are 9346 1451 173 given in the form of diagonal matrix of uncertainties for 6-th 9346 1451 174 neutron energy intervals (LB=1) 9346 1451 175 In the energy range 0.1 - 20 MeV uncertainties are presented 9346 1451 176 in the form of relative covariance matrix for the 48 -neutron 9346 1451 177 energy groups (LB=5). Eigenvalues of this 6-th digits relative 9346 1451 178 covariance matrix are the following: 9346 1451 179 9346 1451 180 1.49939E-07 1.52531E-07 1.53109E-07 1.53318E-07 9346 1451 181 1.54049E-07 1.55336E-07 1.56641E-07 1.58051E-07 9346 1451 182 1.58347E-07 1.60763E-07 1.63743E-07 1.64776E-07 9346 1451 183 1.67513E-07 1.69820E-07 1.73330E-07 1.75751E-07 9346 1451 184 1.77557E-07 2.05370E-07 2.66364E-07 2.70615E-07 9346 1451 185 2.75409E-07 2.80956E-07 2.93138E-07 3.15582E-07 9346 1451 186 3.35128E-07 4.86134E-07 7.89097E-07 3.07453E-06 9346 1451 187 1.65989E-05 9.67332E-05 1.24983E-04 1.51845E-04 9346 1451 188 4.05040E-04 4.14092E-04 5.11273E-04 5.63190E-04 9346 1451 189 7.24776E-04 8.29295E-04 9.68770E-04 9.77438E-04 9346 1451 190 1.07709E-03 1.13969E-03 1.57856E-03 1.84372E-03 9346 1451 191 2.04068E-03 2.65362E-03 3.89983E-03 1.60049E-02 9346 1451 192 9346 1451 193 References: 9346 1451 194 1. S.F.Mughabghab Neutron Cross Sections. N.Y.-London, Academic 9346 1451 195 Press, 1984, v.1, part B. 9346 1451 196 2. J.Kimura Nucl. Sci. Techn., v. 30, p.863, 1993 9346 1451 197 3. A.D.Carlson et al. Proc. of an Int. Conf. on Nuclear Data for 9346 1451 198 Science and Technology, Gatlinburg, Tennessee, USA, 9-13 May 9346 1451 199 1994, Ed. J.Dickens, ANS, Oak Ridge, 1994, v.1, p.40 9346 1451 200 4. A.A.Goverdovskiy et al. Jadernaja Fizika (Rus.), v.58, p.27, 9346 1451 201 January 1995 9346 1451 202 5. Y.Kikuchi Proc. of an Int. Conf. on Nuclear Data for Science 9346 1451 203 and Technology, Gatlinburg, Tennessee, USA, 9-13 May 1994, 9346 1451 204 Ed. J.Dickens, ANS, Oak Ridge, 1994, v.2, p.692 9346 1451 205 6. P.Young, E.Arthur, F.Mann ENDF/B-VI library, MAT 9346, 1990 9346 1451 206 7. E.D.Klema Phys. Rev., v.72, p.88, July 1947 9346 1451 207 8. R.L.Henkel Report LA-2114, 1957 9346 1451 208 R.L.Henkel Report LA-2122, June 1957 9346 1451 209 9. A.N.Protopopov et al. Atomnaja Energija (Sov.),v.4,p.190,1958 9346 1451 210 10. H.W.Schmitt, R.B.Murrey Phys. Rev., v.116, p.1575, 1959 9346 1451 211 10. R.B.Murrey, H.W.Schmitt Phys. Rev., v.115, p.1707, 1959 9346 1451 212 11. B.M.Gokhberg et al. Doklady Akademii Nauk SSSR, v.128,(6), 9346 1451 213 p.1157, 1959 9346 1451 214 12. V.M.Pankratov et al. Atomnaya Energiya, v.9, p.399, 1960 ??? 9346 1451 215 13. V.M.Pankratov Atomnaja Energiya (Sov.), v.14, p.177, 1963 9346 1451 216 14. J.A.Grundl Nucl. Sci. Eng., v.30, p.39, October 1967 9346 1451 217 15. P.H.White, G.P.Warner J. of Nucl. Energy, v.21, p.671, 1967 9346 1451 218 16. W.E.Stein, R.K.Smith, H.L.Smith Proceedings Conf. Neutron 9346 1451 219 Cross Sections and Technology, Washington, D.C., March 4-7, 9346 1451 220 1968, NBS Special Publication 299 , p.627 , U.S. National 9346 1451 221 Bureau of Standards, 1968 9346 1451 222 17. R.H.Iyer, R.Sampathkumar Proc. of 12th Symp. on Nuclear Phys. 9346 1451 223 and Solid State Phys., Roorkee, India, 28-31 December 1969, 9346 1451 224 v.2, p.289 ; 9346 1451 225 R.H.Iyer, R.Sampathkumar Prog. Report BARC/I-79, p.55, 1970 9346 1451 226 18. W.K.Brown et al. Nucl. Phys., vA156, p.609, November 1970 9346 1451 227 19. R.J.Jiacoletti, W.K.Brown, H.G.Olson. Nucl. Sci. Eng., v.48, 9346 1451 228 p.412, August 1972 9346 1451 229 20. V.M.Kupriyanov et al. Atomnaya Energiya, v.45, p.440, 1978 9346 1451 230 21. D.J.Grady et al. Proc. of an Int. Conf. on Nuclear Cross 9346 1451 231 Sections for Technology, Knoxville, Tennessee, 22-26 October 9346 1451 232 1979, NBS Special Publ.594, p.976, September 1980 9346 1451 233 22. R.Arlt et al. Report ZFK-410, p.122, January 1980 ; 9346 1451 234 R.Arlt et al. Kernenergie, v.24, p.48, February 1981 9346 1451 235 23. J.W.Behrens, J.C.Browne, J.C.Malden Nucl. Sci. Eng., v.80, 9346 1451 236 p.393, 1982 9346 1451 237 24. M.Cance, G.Grenier Proc of an Int. Conf. on Nuclear Data 9346 1451 238 for Science and Technology, Antwerp, 6-10 September 1982, 9346 1451 239 Dr. Reidel Publishing Company, 1983, p.51 9346 1451 240 25. M.Varnagy, S.Juhasz, J.Csikai Nucl. Instr. Meth., v.196, 9346 1451 241 p.465, May 1982 9346 1451 242 26. J.W.Meadows Nucl. Sci. Eng., v.85, p.271, November 1983 9346 1451 243 27. I.D.Alkhazov, E.A.Ganza, L.V.Drapchinskij et al. Proceedings 9346 1451 244 of the 3-d All-union Conference on the Neutron Radiation 9346 1451 245 Metrology at Reactors and Accelerators, v.2, p.201, Moscow, 9346 1451 246 CNIIATOMINFORM, 1983 9346 1451 247 28. K.R.Zasadny et al. J. Trans. Amer. Nucl. Soc., v.47, p.425, 9346 1451 248 November 1984 9346 1451 249 29. Wu Jingxia et al. Chinese J. of Nuclear Physics, v.6, p.369, 9346 1451 250 November 1984 9346 1451 251 30. I.Garlea et al. Revue Roumaine de Physique, v.29, p.421, 1984 9346 1451 252 31. A.A.Goverdovski et al. Preprint FEI-1552, Obninsk, 1984 9346 1451 253 32. A.A.Goverdovskij et al. VANT, Serija: Yadernye Konstanty, 9346 1451 254 issue 3(57), p.13, September 1984 9346 1451 255 33. K.Kanda et al. Report JAERI-M-85-035, p.220, 1985 9346 1451 256 34. R.Arlt et al. Isotopenpraxis, v.21, p.344, 1985 9346 1451 257 35. A.A.Goverdovskij et al. Atomnaja Energija (Sov.), v.58, no.2,9346 1451 258 p.137, February 1985 9346 1451 259 36. H.Terayama et al. Progress Report NETU-47, 1986 ; 9346 1451 260 H.Terayama et al. Progress Report NEANDC(J)-122, Sep. 1986 9346 1451 261 37. I.D.Alkhazov et al. VANT, Serija: Yadernye Konstanty, v.4, 9346 1451 262 p.19, December 1986 9346 1451 263 38. V.A.Kalinin et al. VANT, Serija: Yadernye Konstanty, v.4, 9346 1451 264 p.3, December 1987 9346 1451 265 39. P.W.Lisowski et al. Proc. of an Int. Conf. on Nuclear Data 9346 1451 266 for Science and Technology, Mito, Japan, 30 May-3 June 1988, 9346 1451 267 pp. 97-99 9346 1451 268 40. J.W.Meadows Annals of Nucl. Energy, v.15, p.421, August 1988 9346 1451 269 41. J.W.Meadows, D.L.Smith, L.P.Geraldo Annals of Nucl. Energy, 9346 1451 270 v.16, p.471, September 1989 9346 1451 271 36. L.Desdin, S.Szegedi, J.Csikai Acta Physica Hungaria, v.65, 9346 1451 272 p.271, 1989 9346 1451 273 43. P.W.Lisowski et al. Proc. of the Conference on Fifty Years 9346 1451 274 with Nuclear Fission, NIST, Gaithersburg, MD, 1989, p.443 9346 1451 275 44. K.Merla et al. Proc. of an Int. Conf. on Nuclear Data for 9346 1451 276 Science and Technology, Julich, FRG, 13-17 May 1991. Springer 9346 1451 277 Verlag, Berlin - Heidelberg, 1992, p.510-513 9346 1451 278 45. A.D.Carlson et al. Proc. of an Int. Conf. on Nuclear Data for 9346 1451 279 Science and Technology, Gatlinburg, Tennessee, USA, May 9-13, 9346 1451 280 1994, Vol.1, pp. 40-42 9346 1451 281 46. T.Iwasaki et al. Nucl. Sci. Technology, v.36, p.127, 1999 9346 1451 282 47. L.Weston et al. ORNL,LANL eval. April 1989, ENDF/B-VI Library 9346 1451 283 MAT=9228; 9346 1451 284 H.Conde ed. Nuclear Data Standards for Nuclear Measurements, 9346 1451 285 Report NEANDC-311 U, p.51-58, OECD, Paris, 1992. 9346 1451 286 48. M.Kawai et al. eval. NAIG, JENDL-3.2 library, MAT=9437, 9346 1451 287 Rev. 2, February 1993 9346 1451 288 49. P.Young et al. eval. ORNL, LANL , ENDF/B-VI library, MAT=9437,9346 1451 289 Rev.1, January 1993 9346 1451 290 50. S.A.Badikov et al. Preprint FEI-1686, Obninsk, 1985 9346 1451 291 51. S.Badikov, N.Rabotnov, K.Zolotarev Proc. of NEANSC Speciali- 9346 1451 292 st's Meeting on Evaluation and Processing of Covariance Data, 9346 1451 293 Oak Ridge, USA, 1992, OECD, Paris, 1993, p.105 9346 1451 294 52. L.W.Weston et al. Evaluated Neutron Data for Uranium-235, 9346 1451 295 ENDF/B-VI Library, MAT=9228, MF=5, MT=18, eval. April 1989 9346 1451 296 53. W.Mannhart IAEA-TECDOC-410, p.158, IAEA, Vienna, 1987 9346 1451 297 54. W.Mannhart Prog. Report INDC(Ger)-045, pp.40-43, June 1999 9346 1451 298 55. W.Mannhart Validation of Differential Cross Sections with 9346 1451 299 Integral Data , Report INDC(NDS)-435, pp.59-64, IAEA, Vienna, 9346 1451 300 September 2002 9346 1451 301 ******************************************************************9346 1451 302 ----- MF=3 MT= 1 ----- 9346 1451 303 ----- MF=3 MT= 2 ----- 9346 1451 304 93-NP-237 LANL EVAL-APR90 P.YOUNG, E.ARTHUR, F.MANN 9346 1451 305 DIST-SEP91 REV1-JUL91 19910806 9346 1451 306 ----ENDF/B-VI MATERIAL 9346 REVISION 1 9346 1451 307 MT=1 Neutron Total Cross Section. Derrien evaluation (De80, 9346 1451 308 De89) used from 1.0E-5 eV to 0.008 MeV. Above 0.008 MeV,9346 1451 309 coupled channel optical model calculations (ECIS code) 9346 1451 310 with modified Lagrange (Ha80) potential used as prior 9346 1451 311 in a covariance analysis of measurement by Lisowski et 9346 1451 312 al. from 3 to 20 MeV (Li90). 9346 1451 313 MT=2 All Energies, based on subtraction of MT=4,16,17,18,37, 9346 1451 314 and 102 from MT=1. 9346 1451 315 ******************************************************************9346 1451 316 The cross sections for total and elastic were modified between 9346 1451 317 6 and 10 Kev for continuity with the data of Zolatarev file 18. 9346 1451 318 The number of interpolation schemes is set to 1. 9346 1451 319 9346 1451 320 ******************************************************************9346 1451 321 ***************** Program LINEAR (VERSION 2002-1) ***************9346 1451 322 For All Data Greater than 1.0000E-10 barns in Absolute Value 9346 1451 323 Data Linearized to Within an Accuracy of .100000000 per-cent 9346 1451 324 ***************** Program RECENT (VERSION 2002-1) ***************9346 1451 325 for All Data Greater than 1.0000E-10 barns in Absolute Value 9346 1451 326 Data Linearized to within an Accuracy of .100000000 per-cent 9346 1451 327 ***************** Program SIGMA1 (VERSION 2002-1) ***************9346 1451 328 Data Doppler Broadened to 300.000000 Kelvin 9346 1451 329 for All Data Greater than 1.0000E-10 barns in Absolute Value 9346 1451 330 Data Linearized to Within an Accuracy pf .100000000 per-cent 9346 1451 331 ***************** Program FIXUP (Version 2002-1) ****************9346 1451 332 Corrected ZA/AWR in All Sections-----------------------------Yes 9346 1451 333 Corrected Thresholds-----------------------------------------Yes 9346 1451 334 Extended Cross Sections to 20 MeV----------------------------No 9346 1451 335 Allow Cross Section Deletion---------------------------------No 9346 1451 336 Allow Cross Section Reconstruction---------------------------No 9346 1451 337 Make All Cross Sections Non-Negative-------------------------Yes 9346 1451 338 Delete Energies Not in Ascending Order-----------------------Yes 9346 1451 339 Deleted Duplicate Points-------------------------------------Yes 9346 1451 340 Check for Ascending MAT/MF/MT Order--------------------------Yes 9346 1451 341 Check for Legal MF/MT Numbers--------------------------------Yes 9346 1451 342 Allow Creation of Missing Sections---------------------------No 9346 1451 343 Allow Insertion of Energy Points-----------------------------No 9346 1451 344 Create Uniform Energy Grid-----------------------------------No 9346 1451 345 Delete Section if Cross Section =0 at All Energies-----------Yes 9346 1451 346 ***************** Program GROUPIE (VERSION 2002-1) **************9346 1451 347 Unshielded Group Averages Using 640 Groups 9346 1451 348 Weighting Spectrum: Flat (Constant) Spectrum 9346 1451 349 1 451 353 19346 1451 350 2 151 4 19346 1451 351 3 18 217 19346 1451 352 33 18 220 19346 1451 353 9346 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1.75519- 4 1.25445- 4 1.03707- 4 1.13744- 4 1.49423- 4934633 18 141 2.01619- 4 2.62310- 4 3.25822- 4 3.88640- 4 4.77013- 4 5.82359- 4934633 18 142 6.47331- 4 6.28928- 4 4.68876- 4 2.80692- 4 2.74535- 4 3.76575- 4934633 18 143 3.79260- 4 3.00839- 4 2.37506- 4 2.23378- 4 2.48346- 4 2.91552- 4934633 18 144 3.34110- 4 3.62044- 4 3.67635- 4 3.50775- 4 3.17798- 4 2.75835- 4934633 18 145 2.28270- 4 1.78541- 4 1.35636- 4 1.10939- 4 1.11186- 4 1.36234- 4934633 18 146 1.81486- 4 2.41036- 4 3.09547- 4 3.82894- 4 4.95670- 4 6.43341- 4934633 18 147 7.07560- 4 6.18587- 4 3.85226- 4 2.53712- 4 3.15080- 4 3.68519- 4934633 18 148 3.32449- 4 2.69756- 4 2.32252- 4 2.31316- 4 2.58765- 4 3.00781- 4934633 18 149 3.42482- 4 3.69355- 4 3.69506- 4 3.38026- 4 2.81091- 4 2.14757- 4934633 18 150 1.57354- 4 1.21495- 4 1.11377- 4 1.24852- 4 1.56848- 4 2.01899- 4934633 18 151 2.55368- 4 3.13766- 4 3.74637- 4 4.66943- 4 5.87737- 4 7.05021- 4934633 18 152 6.00325- 4 3.65527- 4 2.45354- 4 2.69836- 4 3.08775- 4 3.08003- 4934633 18 153 2.81394- 4 2.52772- 4 2.38472- 4 2.47291- 4 2.80683- 4 3.29499- 4934633 18 154 3.70382- 4 3.71371- 4 3.14563- 4 2.19576- 4 1.33640- 4 9.32197- 5934633 18 155 1.02809- 4 1.45576- 4 2.01834- 4 2.58151- 4 3.07952- 4 3.49183- 4934633 18 156 3.82085- 4 4.17720- 4 4.48871- 4 7.68546- 4 6.41955- 4 3.49689- 4934633 18 157 2.04361- 4 2.22276- 4 2.79420- 4 3.09615- 4 3.05936- 4 2.84298- 4934633 18 158 2.65430- 4 2.67335- 4 2.97649- 4 3.43520- 4 3.68953- 4 3.36988- 4934633 18 159 2.48764- 4 1.51052- 4 9.46217- 5 9.63766- 5 1.41923- 4 2.07376- 4934633 18 160 2.73942- 4 3.31237- 4 3.75312- 4 4.05894- 4 4.28578- 4 4.29956- 4934633 18 161 8.20572- 4 6.63832- 4 3.85859- 4 2.29267- 4 2.05857- 4 2.44471- 4934633 18 162 2.93239- 4 3.28924- 4 3.45667- 4 3.47058- 4 3.41336- 4 3.35468- 4934633 18 163 3.27203- 4 3.03216- 4 2.53079- 4 1.86818- 4 1.30567- 4 1.05473- 4934633 18 164 1.16279- 4 1.55568- 4 2.12285- 4 2.76912- 4 3.42950- 4 4.06580- 4934633 18 165 4.92853- 4 5.90847- 4 8.28554- 4 6.74958- 4 4.22994- 4 2.53188- 4934633 18 166 1.92841- 4 2.07941- 4 2.60164- 4 3.19879- 4 3.65641- 4 3.83410- 4934633 18 167 3.68231- 4 3.25922- 4 2.69812- 4 2.12901- 4 1.63729- 4 1.28584- 4934633 18 168 1.12792- 4 1.18833- 4 1.45362- 4 1.88447- 4 2.43400- 4 3.05984- 4934633 18 169 3.72865- 4 4.76014- 4 6.11963- 4 7.49516- 4 6.08925- 4 4.15996- 4934633 18 170 2.67354- 4 1.87169- 4 1.69933- 4 2.01766- 4 2.64388- 4 3.32589- 4934633 18 171 3.73902- 4 3.59443- 4 2.86151- 4 1.87900- 4 1.12294- 4 8.49159- 5934633 18 172 1.00836- 4 1.41343- 4 1.89882- 4 2.37043- 4 2.79133- 4 3.15565- 4934633 18 173 3.47023- 4 3.86721- 4 4.31509- 4 6.36549- 4 5.39459- 4 3.99082- 4934633 18 174 2.69959- 4 1.81145- 4 1.47503- 4 1.72772- 4 2.43875- 4 3.23360- 4934633 18 175 3.56628- 4 3.08074- 4 2.00343- 4 1.00419- 4 5.98365- 5 8.23513- 5934633 18 176 1.41452- 4 2.08688- 4 2.66476- 4 3.07850- 4 3.32256- 4 3.41989- 4934633 18 177 3.34886- 4 3.02380- 4 5.57432- 4 4.93285- 4 3.90762- 4 2.86104- 4934633 18 178 2.07425- 4 1.75261- 4 1.96933- 4 2.55178- 4 3.04268- 4 2.95024- 4934633 18 179 2.20720- 4 1.28196- 4 7.22802- 5 7.26645- 5 1.15428- 4 1.76292- 4934633 18 180 2.36288- 4 2.85266- 4 3.19732- 4 3.39905- 4 3.46390- 4 3.26467- 4934633 18 181 5.20814- 4 4.90773- 4 4.23620- 4 3.40991- 4 2.65742- 4 2.19526- 4934633 18 182 2.13609- 4 2.35429- 4 2.48687- 4 2.22352- 4 1.62363- 4 1.03524- 4934633 18 183 7.51805- 5 8.40554- 5 1.20915- 4 1.72526- 4 2.28244- 4 2.81467- 4934633 18 184 3.28873- 4 3.85829- 4 4.39684- 4 5.44209- 4 5.44137- 4 4.94518- 4934633 18 185 4.07637- 4 3.07249- 4 2.25930- 4 1.89121- 4 1.91727- 4 1.97704- 4934633 18 186 1.76050- 4 1.30261- 4 8.65519- 5 6.68813- 5 7.80470- 5 1.15832- 4934633 18 187 1.72057- 4 2.38893- 4 3.10384- 4 4.17586- 4 5.51601- 4 6.17393- 4934633 18 188 6.25532- 4 5.63821- 4 4.44969- 4 3.05856- 4 1.97860- 4 1.52979- 4934633 18 189 1.54588- 4 1.57496- 4 1.36506- 4 1.01383- 4 7.50123- 5 7.28036- 5934633 18 190 9.84158- 5 1.48103- 4 2.15526- 4 2.94497- 4 4.24091- 4 6.00908- 4934633 18 191 7.01848- 4 6.99402- 4 6.10456- 4 4.54759- 4 2.85268- 4 1.63429- 4934633 18 192 1.13789- 4 1.10696- 4 1.15827- 4 1.13531- 4 1.09834- 4 1.15820- 4934633 18 193 1.37931- 4 1.77038- 4 2.30735- 4 2.95477- 4 4.06168- 4 5.64219- 4934633 18 194 7.79050- 4 7.69067- 4 6.54107- 4 4.58342- 4 2.51297- 4 1.08549- 4934633 18 195 5.80237- 5 7.41570- 5 1.16655- 4 1.60384- 4 1.97857- 4 2.30396- 4934633 18 196 2.61252- 4 2.92868- 4 3.26450- 4 3.81206- 4 4.60124- 4 8.69495- 4934633 18 197 8.52683- 4 6.90019- 4 4.27097- 4 1.77648- 4 4.11395- 5 3.38196- 5934633 18 198 1.06327- 4 2.00867- 4 2.82548- 4 3.39481- 4 3.72530- 4 3.87081- 4934633 18 199 3.88907- 4 3.77309- 4 3.51727- 4 9.64193- 4 9.01880- 4 6.57344- 4934633 18 200 3.41072- 4 1.07066- 4 3.11372- 5 8.47740- 5 1.96925- 4 3.10075- 4934633 18 201 3.95536- 4 4.45941- 4 4.64516- 4 4.58009- 4 4.14746- 4 3.25813- 4934633 18 202 9.79517- 4 8.48395- 4 5.60783- 4 2.68204- 4 9.54045- 5 6.53598- 5934633 18 203 1.30517- 4 2.32221- 4 3.30598- 4 4.06862- 4 4.56004- 4 4.79820- 4934633 18 204 4.75834- 4 4.25947- 4 8.85772- 4 7.28861- 4 4.60726- 4 2.13428- 4934633 18 205 6.81451- 5 3.16448- 5 7.13005- 5 1.50030- 4 2.40988- 4 3.28775- 4934633 18 206 4.06173- 4 4.96655- 4 5.78294- 4 7.35670- 4 5.68859- 4 3.21582- 4934633 18 207 1.03067- 4-2.84224- 5-6.78220- 5-3.62829- 5 4.05124- 5 1.41735- 4934633 18 208 2.53158- 4 4.20532- 4 6.26698- 4 5.22812- 4 3.55656- 4 1.57839- 4934633 18 209 2.11845- 6-8.22425- 5-9.78357- 5-6.04168- 5 1.27671- 5 1.07588- 4934633 18 210 2.69172- 4 4.90480- 4 3.05310- 4 2.08198- 4 1.10496- 4 3.95566- 5934633 18 211 3.42473- 6-1.04504- 6 1.90253- 5 5.63655- 5 1.32607- 4 2.49668- 4934633 18 212 2.37424- 4 2.42348- 4 2.29411- 4 2.06211- 4 1.78849- 4 1.51149- 4934633 18 213 1.25140- 4 9.14376- 5 5.55298- 5 3.49455- 4 4.09754- 4 4.23490- 4934633 18 214 4.00001- 4 3.50544- 4 2.84715- 4 1.70068- 4 9.99255- 6 5.36145- 4934633 18 215 5.97068- 4 6.00550- 4 5.60150- 4 4.89289- 4 3.47930- 4 1.33595- 4934633 18 216 7.05793- 4 7.51272- 4 7.45658- 4 7.02138- 4 5.88980- 4 3.96187- 4934633 18 217 8.46993- 4 8.93652- 4 9.01817- 4 8.60734- 4 7.52757- 4 1.00335- 3934633 18 218 1.07891- 3 1.14035- 3 1.16290- 3 1.23137- 3 1.41398- 3 1.59674- 3934633 18 219 1.79587- 3 2.24847- 3 3.07560- 3 934633 18 220 934633 099999 9346 0 0 0 0 0 0 0 9.42390E+4 2.36999E+2 0 1 34 109437 1451 1 0.0 1.00000E+0 0 0 0 69437 1451 2 1.00000E+0 2.00000E+7 0 0 10 20029437 1451 3 3.00000E+2 0.0 1 0 564 49437 1451 4 94-Pu-239 NAIG EVAL-MAR87 M.KAWAI, T.YOSHIDA, K.HIDA 9437 1451 5 DIST-Feb2004 9437 1451 6 ----IRDF-2002 MATERIAL 9437 9437 1451 7 -----INCIDENT NEUTRON DATA 9437 1451 8 ------ENDF-6 FORMAT 9437 1451 9 ================================================================= 9437 1451 10 94-PU-239 NAIG EVAL-MAR87 M.KAWAI, T.YOSHIDA, K.HIDA 9437 1451 11 DIST-SEP89 REV2-FEB93 9437 1451 12 ----JENDL-3.2 MATERIAL 9437 9437 1451 13 ================================================================= 9437 1451 14 HISTORY 9437 1451 15 87-03 EVALUATION WAS MADE BY 9437 1451 16 M.KAWAI AND K.HIDA(NAIG) : CROSS SECTIONS ABOVE 9437 1451 17 RESONANCE REGION AND OTHER QUANTITIES, 9437 1451 18 T.YOSHIDA(NAIG) : RESONANCE PARAMETERS AND BACKGROUND 9437 1451 19 CROSS SECTIONS, 9437 1451 20 DATA WERE COMPILED BY T.NAKAGAWA (JAERI). 9437 1451 21 88-08 PARTLY MODIFIED. 9437 1451 22 NU-BAR, RESOLVED RESONS., (N,2N). 9437 1451 23 89-02 FP YIELDS WERE TAKEN FROM JNDC FP DECAY DATA FILE VERSION-2.9437 1451 24 89-03 UNRESOLVED RESONANCE PARAMETERS WERE SLIGHTLY MODIFIED. 9437 1451 25 93-02 JENDL-3.2. 9437 1451 26 RESONANCE PARAMETERS EVALUATED BY H.DERRIEN (JAERI)/1/. 9437 1451 27 FISSION SPECTRA CALCULATED BY T.OHSAWA (KINKI UNIV.) 9437 1451 28 COMPILED BY T.NAKAGAWA (NDC/JAERI) 9437 1451 29 9437 1451 30 ***** MODIFIED PARTS FOR JENDL-3.2 ******************** 9437 1451 31 (2,151) RESOLVED RESONANCE PARAMETERS UP TO 2.5 KEV 9437 1451 32 *********************************************************** 9437 1451 33 9437 1451 34 9437 1451 35 MF=1 GENERAL INFORMATION 9437 1451 36 MT=451 DESCRIPTIVE DATA AND DICTIONARY 9437 1451 37 9437 1451 38 MF=2 RESONANCE PARAMETERS 9437 1451 39 MT=151 RESOLVED AND UNRESOLVED RESONANCE PARAMETERS 9437 1451 40 RESOLVED RES. PARAMETERS FOR REICH-MOORE FORMULA: UP TO 2.5 KEV9437 1451 41 PARAMETERS WERE TAKEN FROM REF./1/ DETAILS OF EVALUATION 9437 1451 42 ARE GIVEN IN APPENDIX. 9437 1451 43 UNRESOLVED RESONANCES : FROM 2.5 TO 30 KEV. 9437 1451 44 THE ENERGY DEPENDENT S0, S1 AND FISSION WIDTH WERE DETER- 9437 1451 45 MINED SO AS TO REPRODUCE THE EVALUATED TOTAL, CAPTURE AND 9437 1451 46 FISSION CROSS SECTIONS. 9437 1451 47 9437 1451 48 2200-M/SEC CROSS SECTIONS AND CALCULATED RESONANCE INTEGRALS. 9437 1451 49 2200 M/S RES. INTEG. 9437 1451 50 TOTAL 1025.7 B - 9437 1451 51 ELASTIC 7.970 B - 9437 1451 52 FISSION 747.4 B 302.6 B 9437 1451 53 CAPTURE 270.3 B 181.6 B 9437 1451 54 9437 1451 55 MF=3 NEUTRON CROSS SECTIONS 9437 1451 56 BELOW 2.5 KEV, CROSS SECTIONS WERE REPRESENTED WITH THE 9437 1451 57 RESOLVED RESONANCE PARAMETERS. BETWEEN 2.5 AND 30 KEV, CROSS9437 1451 58 SECTIONS WERE REPLACED WITH UNRESOLVED RESONANCE PARAMETERS. 9437 1451 59 9437 1451 60 MT=1 TOTAL 9437 1451 61 BELOW 7 MEV, JENDL-2 EVALUATION WHICH WERE BASED ON THE 9437 1451 62 EXPERIMENTS OF REFS./10,11,12,13,14/ WAS ADOPTED. ABOVE 9437 1451 63 7 MEV, EXPERIMENTAL DATA BY POENITZ /15/ WERE ADOPTED. 9437 1451 64 9437 1451 65 MT=2 ELASTIC SCATTERING 9437 1451 66 CALCULATED AS (TOTAL) - (PARTIAL CROSS SECTIONS). 9437 1451 67 9437 1451 68 MT=18 FISSION 9437 1451 69 BELOW 50 KEV 9437 1451 70 BASED ON MEASUREMENTS OF REF./31/ AND REF./32/. 9437 1451 71 ABOVE 50 KEV 9437 1451 72 SIMULTANEOUS EVALUATION WAS PERFORMED BY KANDA ET AL./33/ 9437 1451 73 9437 1451 74 MF=33 COVARIANCES 9437 1451 75 MT=18 9437 1451 76 TAKEN FROM JENDL/D-99. 9437 1451 77 9437 1451 78 REFERENCES 9437 1451 79 1) DERRIEN H.: J. NUCL. SCI. TECHNOL., 30, 845 (1993). 9437 1451 80 10) UTTELY C.A.: EANDC(UK)-40 (1964). 9437 1451 81 11) SCHWARTZ R.B. ET AL.: NUCL. SCI. ENG., 54, 322 (1974). 9437 1451 82 12) FOSTER D.G.JR. AND GLAGOW D.W.: PHYS. REV., C3, 576 (1971). 9437 1451 83 13) SMITH A.B. ET AL.: J. NUCL. ENERGY, 27, 317 (1973). 9437 1451 84 14) NADOLNY ET AL.: C00-3058-39, 33 (1973). 9437 1451 85 15) POENITZ W.P. ET AL.: NUCL. SCI. ENG., 78, 333 (1981). 9437 1451 86 31) GAYTHER D.B.: 1975 WASHINGTON, 2, 560 (1975). 9437 1451 87 32) WAGEMANS C. ET AL.: ANN. NUCL. ENERGY, 7, 495 (1980). 9437 1451 88 9437 1451 89 9437 1451 90 ================================================================= 9437 1451 91 APPENDIX RESONANCE DATA 9437 1451 92 ================================================================= 9437 1451 93 9437 1451 94 THE PRESENT FILE CONTAINS THE RESONANCE PARAMETERS OBTAINED 9437 1451 95 FROM A SAMMY FIT ANALYSIS OF HIGH RESOLUTION EXPERIMENTAL DATA, 9437 1451 96 PERFORMED AT ORNL(OAK RIDGE NATIONNAL LABORATORY,USA) BY H.DERRIEN9437 1451 97 AND G.DE SAUSSURE AND AT JAERI(TOKAI-MURA RESEARCH ESTABLISHMENT, 9437 1451 98 JAPAN) BY H.DERRIEN. 9437 1451 99 THE FILE CONTAINS THREE INDEPENDANT SECTIONS: 9437 1451 100 1/ THE FIRST CORRESPONDS TO THE ENERGY RANGE 0 KEV TO 1 KEV. 9437 1451 101 THE CORRESPONDING SET OF RESONANCE PARAMETRES CONTAINS 398 RESO- 9437 1451 102 NANCES IN THE ENERGY RANGE 0 KEV TO 1 KEV, 4 FICTICIOUS NEGATIVE 9437 1451 103 ENERGY RESONANCES AND 3 FICTICIOUS RESONANCES ABOVE 1 KEV; 9437 1451 104 2/ THE SECOND CORRESPONDS TO THE ENERGY RANGE 1 KEV TO 2 KEV. 9437 1451 105 THE CORRESPONDING SET OF RESONANCE PARAMETERS CONTAINS 435 RESON- 9437 1451 106 ANCES IN THE ENERGY RANGE 0.980 KEV TO 2.02 KEV, 3 FICTICIOUS 9437 1451 107 RESONANCES BELOW 0.9 KEV AND 3 FICTICIOUS RESONANCES ABOVE 2.02 9437 1451 108 KEV; 9437 1451 109 3/ THE THIRD CORRESPONDS TO THE ENERGY RANGE 2 KEV TO 2.5 KEV. 9437 1451 110 THE CORRESPONDING SET OF RESONANCE PARAMETERS CONTAINS 218 RESO- 9437 1451 111 NANCES IN THE ENERGY RANGE 1.98 KEV TO 2.53 KEV, 3 FICTICIOUS 9437 1451 112 RESONANCES BELOW 1.98 KEV AND 3 FICTICIOUS RESONANCES ABOVE 2.53 9437 1451 113 KEV. 9437 1451 114 IN ALL SECTIONS THE FICTICIOUS RESONANCE PARAMETERS TAKE INTO 9437 1451 115 ACCOUNT THE CONTRIBUTION OF ALL THE EXTERNAL TRUNCATED RESONANCES 9437 1451 116 IN SUCH A WAY THAT NO TOTAL, SCATTERING, FISSION AND CAPTURE 9437 1451 117 SMOOTH FILES ARE NEEDED IN THE CORRESPONDING ENERGY RANGES FOR THE9437 1451 118 REPRODUCTION OF THE CROSS SECTIONS WITHIN THE EXPERIMENTAL ERRORS.9437 1451 119 THE FOLLOWING EXPERIMENTAL DATA BASE HAS BEEN USED IN THE SAMMY9437 1451 120 FITS: 9437 1451 121 -ABSORPTION AND FISSION FROM R.GWIN ET AL./1,2/; 9437 1451 122 -FISSION FROM R.GWIN ET AL./3,4/, J.BLONS/5/, L.W.WESTON ET 9437 1451 123 AL./6,7/; 9437 1451 124 -TRANSMISSION FROM R.R.SPENCER ET AL./8/, J.A.HARVEY ET AL./9/ 9437 1451 125 PRIOR TO THE FITS THE EXPERIMENTAL FISSION AND ABSORPTION CROSS 9437 1451 126 SECTIONS WERE NORMALISED, DIRECTLY OR INDIRECTLY TO THE 0.0253 EV 9437 1451 127 VALUES OBTAINED BY THE ENDF/B-VI STANDARD EVALUATION GROUP/10/. 9437 1451 128 THE TRANSMISSION DATA WERE CONSIDERED AS ACCURATE ABSOLUTE MEASU- 9437 1451 129 REMENTS(R.R.SPENCER TOTAL CROSS SECTION AT 0.0253 EV IS 1025.0 B 9437 1451 130 IN EXCELLENT AGREEMENT WITH THE 1027.3 B STANDARD VALUE). 9437 1451 131 DETAILS ON THE ANALYSIS ARE FOUND IN REFERENCES/11,12,13/ 9437 1451 132 9437 1451 133 ***************************************************************** 9437 1451 134 COMMENTS ON THE THERMAL AND LOW ENERGY RANGES 9437 1451 135 ***************************************************************** 9437 1451 136 9437 1451 137 THE THERMAL CROSS SECTION VALUES CALCULATED AT 293 K BY THE 9437 1451 138 RESONANCE PARAMETERS OF THE FIRST SECTION ARE GIVEN IN THE FOLLO- 9437 1451 139 WING TABLE: 9437 1451 140 9437 1451 141 9437 1451 142 SAMMY RESENDD PROPOSED 9437 1451 143 293 K (BARN) STANDARD VALUES(BARN)/10/ 9437 1451 144 **************************************************************** 9437 1451 145 FISSION 747.64 747.90 747.99+-1.87 9437 1451 146 CAPTURE 271.10 270.73 271.43+-2.14 9437 1451 147 SCATTERING 7.97 7.99 7.88+-0.97 9437 1451 148 TOTAL 1026.71 1026.62 1027.30+-5.00 9437 1451 149 **************************************************************** 9437 1451 150 9437 1451 151 ONE SHOULD NOTE THAT THE 293 K CROSS SECTIONS CALCULATED AT 9437 1451 152 0.0253 EV DEPEND ON THE WAY THE DOPPLER BROADENING CALCULATION IS 9437 1451 153 PERFORMED. FOR INSTANCE USING A GAUSSIAN BROADENING FUNCTION WILL 9437 1451 154 GIVE A FISSION CROSS SECTION ABOUT 2.5 BARNS LARGER THAN THE ONE 9437 1451 155 OBTAINED FROM THE ACCURATE CALCULATION WHICH CONSERVES THE 1/V 9437 1451 156 SHAPE OF THE THERMAL CROSS SECTION. THE VALUES GIVEN IN THE TABLE 9437 1451 157 ABOVE WERE OBTAINED FROM SAMMY (LEAL-HWANG METHOD)/14,15/ AND FROM9437 1451 158 RESENDD WITH 0.1% FOR THE INTERPOLATION ACCURACY/16/. 9437 1451 159 THE FOLLOWING TABLE SHOWS EXPERIMENTAL CROSS SECTIONS AVE- 9437 1451 160 RAGED OVER THE ENERGY RANGES 0.02 EV TO 0.06 EV AND 0.02 EV TO 9437 1451 161 0.65 EV, COMPARED TO THE CALCULATED VALUES: 9437 1451 162 9437 1451 163 9437 1451 164 AVERAGE CROSS-SECTIONS(BARN) 9437 1451 165 **************************************************************** 9437 1451 166 REFERENCES(1-10) 0.02 TO 0.06 EV 0.02 TO 0.65 EV 9437 1451 167 **************************************************************** 9437 1451 168 EXP CALC (293K) EXP CALC (293K) 9437 1451 169 GWIN71 FISS 631.41 843.71 9437 1451 170 GWIN76 FISS 631.41 838.39 9437 1451 171 GWIN84 FISS(*) 631.41 631.75(+0.05%) 837.18 838.69(+0.18%) 9437 1451 172 DERUYTER70 FISS 631.41 859.43 9437 1451 173 WAGEMANS80 FISS 631.41 862.56 9437 1451 174 WAGEMANS88 FISS 631.41 841.80 9437 1451 175 GWIN71 CAPTURE 243.84 243.22(-0.25%) 524.75 518.13(-1.26%) 9437 1451 176 GWIN76 ABSORPT(*) 875.90 874.29(-0.18%) 1359.96 1357.14(-0.21%) 9437 1451 177 SPENCER84 TOT(*) 883.20 882.86(-0.04%) 1361.69 1367.6 (+0.43%) 9437 1451 178 **************************************************************** 9437 1451 179 (*)THESE DATA HAD THE LARGIEST WEIGHT IN THE THERMAL FIT. THE VA- 9437 1451 180 LUES BETWEEN THE PARENTHESES GIVE THE PERCENTAGE DEVIATION BETWEEN9437 1451 181 THE CALCULATED DATA AND THE EXPERIMENTAL DATA. 9437 1451 182 ***************************************************************** 9437 1451 183 9437 1451 184 THE VALUE OF 631.4 BARNS FOR ALL THE AVERAGED EXPERIMENTAL 9437 1451 185 FISSION CROSS SECTIONS IN THE ENERGY RANGE 0.02 EV TO 0.06 EV 9437 1451 186 CORRESPONDS TO THE RENORMALISATION OF THE FISSION EXPERIMENTS TO 9437 1451 187 748.0+-1. BARNS AT 0.0253 EV. ORNL DATA ARE CONSISTENT WITHIN 0.8%9437 1451 188 OVER THE ENERGY RANGE 0.02 EV TO 0.65 EV (I.E. OVER THE 0.3 EV 9437 1451 189 RESONANCE). DERUYTER 1970 AND WAGEMANS 1980 DATA ARE ABOUT 2.5% 9437 1451 190 LARGER AND WERE NOT INCLUDED IN THE SAMMY FIT. WHEN NORMALIZED ON 9437 1451 191 THE STANDARD VALUE AT 0.0253 EV, GWIN 76 ABSORPTION AGREES WITH 9437 1451 192 THE ABSORPTION OBTAINED FROM SPENCER TOTAL CROSS SECTION WITHIN 9437 1451 193 0.7% OVER THE 0.3 EV RESONANCE. THE PRESENT EVALUATION IS 9437 1451 194 ESSENTIALLY THE RESULT OF A CONSISTENT SAMMY ANALYSIS OF ALL THE 9437 1451 195 AVAILABLE ORNL DATA WITH A LARGER WEIGHT ON GWIN 1984 FISSION, 9437 1451 196 GWIN 1976 ABSORPTION AND SPENCER TRANSMISSION DATA. 9437 1451 197 AFTER RENORMALIZATION OF THE CALCULATED FISSION CROSS SECTION9437 1451 198 ON THE PRELIMINARY 1991 WESTON AND TODD FISSION DATA(SEE NEXT 9437 1451 199 SECTION) A SLIGHT ADJUSTMENT OF THE NEGATIVE RESONANCE PARAMETERS 9437 1451 200 WAS PERFORMED TO KEEP THE VALUES CALCULATED AT 0.0253 EV IN CLOSE 9437 1451 201 AGREEMENT WITH THE STANDARD VALUES. THE 1988 DATA OF WAGEMANS ET 9437 1451 202 AL./17/ AGREE WITHIN 0.4% WITH THE CALCULATED VALUES OVER THE 9437 1451 203 ENERGY RANGE FROM 0.02 EV TO 0.65 EV AFTER ADJUSTMENT OF THE 9437 1451 204 ENERGY SCALE TO THE ORNL SCALE (THE DIFFERENCE WAS 0.27 EV AT 20 9437 1451 205 EV BETWEEN 1988 WAGEMANS AND ORNL SAMMY FIT ENERGY SCALES). 9437 1451 206 9437 1451 207 ***************************************************************** 9437 1451 208 COMMENTS ON THE 0 KEV TO 1 KEV ENERGY RANGE. 9437 1451 209 ***************************************************************** 9437 1451 210 9437 1451 211 AT THE END OF 1987, AN ANALYSIS WAS COMPLETED UP TO 1 KEV. IN A9437 1451 212 PRELIMINARY STEP, A CORRELATED FIT OF HARVEY TRANSMISSION DATA, 9437 1451 213 WESTON 84 FISSION DATA AND BLONS FISSION DATA WAS PERFORMED, WITH 9437 1451 214 POSSIBLE ADJUSTMENT OF THE NORMALIZATION COEFFICIENTS AND OF THE 9437 1451 215 BACKGROUND CORRECTIONS. THIS PRELIMINARY STEP HAS SHOWN THAT THIS 9437 1451 216 ADJUSTMENT WAS NOT NECESSARY TO HAVE CONSISTENCY BETWEEN HARVEY 9437 1451 217 DATA AND WESTON DATA. BLONS DATA NEEDED A LARGE READJUSTMENT OF 9437 1451 218 THE BACKGROUND AND OF THE NORMALIZATION. THEREFORE, THE FINAL FIT 9437 1451 219 WAS PERFORMED ONLY ON HARVEY TRANSMISSION DATA, GWIN 84 FISSION 9437 1451 220 DATA (BELOW 30 EV) AND WESTON 84 FISSION DATA, WITH NO BACKGROUND 9437 1451 221 AND NORMALISATION ADJUSTMENT. BLONS DATA, WHICH HAVE BETTER 9437 1451 222 RESOLUTION THAN WESTON 84 DATA, WERE USED ONLY TO OBTAIN MORE 9437 1451 223 ACCURATE FISSION WIDTHS OF SOME NARROW RESONANCES IN THE HIGH 9437 1451 224 ENERGY RANGE. 9437 1451 225 IN 1989, PRELIMINARY RESULTS OF THE 1988 WESTON FISSION 9437 1451 226 MEASUREMENT/7/ WERE INCLUDED IN THE SAMMY EXPERIMENTAL DATA 9437 1451 227 BASE. ONE EXPECTED FROM THIS MEASUREMENT, WHICH WAS PERFORMED BY 9437 1451 228 USING A 86 M FLIGHT PATH WITH A RESOLUTION COMPARABLE TO THE 9437 1451 229 RESOLUTION OF HARVEY TRANSMISSION, A CONFIRMATION OF THE EXCELLENT9437 1451 230 QUALITY OF THE 1984 MEASUREMENT. A CONSISTENT SAMMY FIT OF HARVEY 9437 1451 231 TRANSMISSION, WESTON 84 FISSION AND PRELIMINARY WESTON 88 FISSION 9437 1451 232 WAS RESTARTED FROM THE PARAMETER AND COVARIANCE FILES OBTAINED IN 9437 1451 233 1987. IT APPEARED THAT LARGE BACKGROUND AND NORMALISATION 9437 1451 234 CORRECTIONS WERE NEEDED ON THE NEW WESTON FISSION TO OBTAIN 9437 1451 235 CONSISTENCY WITH HARVEY TRANSMISSION DATA. THESE CORRECTIONS WERE 9437 1451 236 COMPARABLE TO THOSE FOUND ON BLONS DATA AND WERE NOT UNDERSTOOD BY9437 1451 237 THE AUTHORS OF THE EXPERIMENT. THE LAST SAMMY RUNS WERE PERFORMED 9437 1451 238 BY NOT ALLOWING BACKGROUND AND NORMALIZATION VARIATIONS ON HARVEY 9437 1451 239 TRANSMISSION AND WESTON 84 FISSION(VERY SMALL ERROR BARS WERE 9437 1451 240 ASSIGNED TO THE CORRESPONDING PARAMETERS IN THE COVARIANCE MATRIX)9437 1451 241 AND BY ALLOWING THESE VARIATIONS ON WESTON 88 DATA. A NEW SET OF 9437 1451 242 RESONANCE PARAMETERS WAS OBTAINED,WHICH WAS IMPROVED COMPARED TO 9437 1451 243 THE PREVIOUS SET DUE TO THE VERY HIGH RESOLUTION OF THE NEW WESTON9437 1451 244 FISSION MEASUREMENT. 9437 1451 245 THE CALCULATED AVERAGE FISSION CROSS SECTION IN THE ENERGY 9437 1451 246 RANGE FROM 0.1 KEV TO 1.0 KEV WAS 3.7% SMALLER THAN THE VALUES 9437 1451 247 OBTAINED BY THE ENDF/B-VI STANDARD EVALUATION GROUP, DUE TO THE 9437 1451 248 FACT THAT WESTON 84 DATA WERE 3.1% LOWER THAN THE AVERAGE STANDARD9437 1451 249 VALUE. A NEW MEASUREMENT WAS PERFORMED BY WESTON AND TODD IN 1991 9437 1451 250 /18/ IN ORDER TO CHECK THEIR 1984 DATA. A CAREFUL NORMALIZATION 9437 1451 251 OF THE DATA IN THE THERMAL ENERGY RANGE SHOWED THAT THE 1984 DATA 9437 1451 252 SHOULD BE RENORMALIZED BY ABOUT +3%. TO TAKE INTO ACCOUNT THIS 9437 1451 253 RENORMALIZATION, THE 1989 RESONANCE PARAMETERS WERE MODIFIED AT 9437 1451 254 JAERI/13/ IN THE FOLLOWING WAY: 9437 1451 255 1/ INCREASE OF THE FISSION WIDTH BY 3% AND DECREASE OF THE 9437 1451 256 CAPTURE WIDTH BY A QUANTITY EQUAL TO THE VARIATION OF THE FISSION 9437 1451 257 WIDTH, IN THE NARROW RESONANCES(MAINLY 1+ RESONANCES); THAT DOES 9437 1451 258 NOT MODIFY THE TOTAL CROSS SECTION IN THE CORRESPONDING 9437 1451 259 RESONANCES; 9437 1451 260 2/ ADJUSTMENT OF THE NEUTRON WIDTH OF THE 0+ RESONANCES BY A 9437 1451 261 REFIT OF THE TRANSMISSION DATA AND OF THE RENORMALIZED WESTON AND 9437 1451 262 TODD 1984 DATA IN ENERGY RANGES WHERE THE CONTRIBUTION OF THE 0+ 9437 1451 263 RESONANCES IS DOMINANT, AND INCREASE OF THE OTHER(SMALL) 0+ 9437 1451 264 NEUTRON WIDTHS BY 3%. NO SEVERE INCONSISTENCY WAS OBSERVED BETWEEN9437 1451 265 THE TRANSMISSION DATA AND THE NEW FISSION DATA OVER THE DOMINANT 9437 1451 266 0+ RESONANCES;THE DIFFERENCES BETWEEN THE 1989 FITS OF THE TRANS- 9437 1451 267 MISSION AND THE NEW FITS WERE CONSISTENT WITHIN THE EXPERIMENTAL 9437 1451 268 ERROR BARS. 9437 1451 269 THE FOLLOWING TABLE SHOWS THE FISSION CROSS SECTIONS CALCULA- 9437 1451 270 TED FROM THE RESONANCE PARAMETERS, THE EXPERIMENTAL VALUES AND THE9437 1451 271 RESULTS OF THE ENDF/B-VI STANDARD EVALUATION GROUP AVERAGED IN THE9437 1451 272 SAME ENERGY INTERVALS. WESTON 1991 DATA ARE PRELIMINARY. WESTON 9437 1451 273 1984 DATA ARE NORMALIZED ON PRELIMINARY WESTON 1991: 9437 1451 274 9437 1451 275 CROSS-SECTIONS(BARN) 9437 1451 276 ************************************************* 9437 1451 277 ENERGY CALCUL WESTON WESTON STANDARD 9437 1451 278 (EV) 1991 1984 9437 1451 279 ************************************************* 9437 1451 280 0.010-10. 80.12 79.98 9437 1451 281 9-20 94.74 94.91 9437 1451 282 20-40 17.52 17.76 17.97 9437 1451 283 40-60 50.64 50.90 50.87 9437 1451 284 60-100 54.42 54.38 54.33 9437 1451 285 100-200 18.63 18.59 18.56 18.66 9437 1451 286 200-300 17.85 17.89 17.88 9437 1451 287 300-400 8.31 8.34 8.43 9437 1451 288 400-500 9.59 9.58 9.57 9437 1451 289 ------------------------------------------------- 9437 1451 290 200-500 11.92 11.93 11.93 11.96 9437 1451 291 ------------------------------------------------- 9437 1451 292 500-600 15.39 15.57 15.86 9437 1451 293 600-700 4.37 4.30 4.46 9437 1451 294 700-800 5.51 5.53 5.63 9437 1451 295 800-900 4.84 4.89 4.98 9437 1451 296 900-1000 8.33 8.38 8.30 9437 1451 297 ------------------------------------------------- 9437 1451 298 500-1000 7.69 7.73 7.73 7.79 9437 1451 299 ------------------------------------------------- 9437 1451 300 20-1000 13.09 13.11 13.11 9437 1451 301 ************************************************** 9437 1451 302 9437 1451 303 9437 1451 304 GWIN 1971 AND 1976 ABSORPTION DATA WERE NOT INCLUDED IN THE 9437 1451 305 SAMMY FIT IN THE ENERGY RANGE ABOVE 1 EV. ACCURATE ABSORPTION 9437 1451 306 CROSS SECTIONS SHOULD BE CALCULATED FROM THE PARAMETERS OBTAINED 9437 1451 307 FROM THE ANALYSIS OF THE TRANSMISSION AND FISSION DATA. THE FOLLO-9437 1451 308 WING TABLE SHOWS THE CALCULATED AVERAGE VALUES OF THE CAPTURE, AB-9437 1451 309 SORPTION AND ALPHA COMPARED TO GWIN 1971 AND GWIN 1976 DATA. THE 9437 1451 310 CALCULATIONS WERE PERFORMED WITH RESENDD, 1.0 % ACCURACY: 9437 1451 311 9437 1451 312 CROSS-SECTIONS(BARN) 9437 1451 313 ******************************************************* 9437 1451 314 ENERGY(EV) CALC. VALUES (293K) GWIN DATA 9437 1451 315 ******************************************************* 9437 1451 316 CAPT ABSORP ALPHA ABSORP ALPHA 9437 1451 317 7.3- 16.0 76.61 196.04 0.64 208.00 0.74(*) 9437 1451 318 16.0- 37.5 20.51 44.55 0.85 46.50 0.89(*) 9437 1451 319 37.5- 50.0 48.72 70.00 2.29 83.15 2.96(*) 9437 1451 320 50.0-100.0 33.60 92.13 0.57 92.84 0.63 9437 1451 321 100.0-200.0 15.58 34.29 0.83 33.66 0.87 9437 1451 322 200.0-300.0 15.85 33.68 0.89 34.69 0.94 9437 1451 323 300.0-400.0 9.69 18.01 1.16 18.31 1.16 9437 1451 324 400.0-500.0 3.96 13.56 0.41 13.56 0.44 9437 1451 325 500.0-600.0 10.87 26.30 0.70 26.54 0.72 9437 1451 326 600.0-700.0 6.53 10.90 1.49 11.57 1.54 9437 1451 327 700.0-800.0 4.95 10.47 0.90 10.52 0.97 9437 1451 328 800.0-900.0 3.65 8.50 0.75 9.30 0.82 9437 1451 329 900.0-999.9 5.06 13.51 0.60 13.23 0.70 9437 1451 330 ****************************************************** 9437 1451 331 (*) GWIN 1971 DATA 9437 1451 332 9437 1451 333 IF ONE EXCEPTS THE ENERGY RANGE 37.5-50 EV, THE CALCULATED AB- 9437 1451 334 SORPTION VALUES AGREE WELL WITH GWIN EXPERIMENTAL DATA; THEY ARE 9437 1451 335 ON AVERAGE 1.2% LOWER IN THE ENERGY RANGE FROM 50 EV TO 1000 EV. 9437 1451 336 9437 1451 337 ***************************************************************** 9437 1451 338 COMMENTS ON THE 1 KEV TO 2 KEV ENERGY RANGE 9437 1451 339 ***************************************************************** 9437 1451 340 9437 1451 341 PRELIMINARY RESONANCE PARAMETERS WERE OBTAINED IN 1989 FROM THE9437 1451 342 ANALYSIS OF THE HARVEY THICK SAMPLE TRANSMISSION DATA AND OF THE 9437 1451 343 PRELIMINARY RESULTS OF WESTON 88 FISSION MEASUREMENT. DUE TO LACK9437 1451 344 OF TIME, THE MEDIUM AND THIN SAMPLE TRANSMISSION DATA WERE NOT 9437 1451 345 INCLUDED IN THE SAMMY DATA BASE, AND THE CONTRIBUTION OF THE 9437 1451 346 TRUNCATED EXTERNAL RESONANCES WAS NOT CAREFULLY INVESTIGATED. 9437 1451 347 NEVERTHELESS, THE RESULTS WERE USED IN THE ENDF/B-VI FILE, ALONG 9437 1451 348 WITH A SMOOTH FILE IN ORDER TO AGREE WITH THE AVERAGE VALUES OF A 9437 1451 349 PREVIOUS ENDF/B-VI EVALUATION (THIS PRELIMINARY SET OF PARAMETERS 9437 1451 350 WAS CONSIDERED AS MORE USEFUL THAN THE STATISTICAL PARAMETERS IN 9437 1451 351 THE ENERGY RANGE 1 KEV TO 2 KEV FOR THE CALCULATION OF THE SELF- 9437 1451 352 SHIELDING FACTORS). 9437 1451 353 THE ANALYSIS WAS RESTARTED IN APRIL 1991 AT JAERI(TOKAI 9437 1451 354 RESEARCH ESTABLISHMENT) WITH AN UPDATED VERSION OF SAMMY ADAPTED 9437 1451 355 BY T.NAKAGAWA TO THE FACOM 780. THE PRELIMINARY SET OF PARAMETERS 9437 1451 356 OBTAINED AT OAK RIDGE IN 1989 WAS USED AS PRIOR INFORMATIONS TO 9437 1451 357 START THE SAMMY CALCULATIONS. ALSO PRIOR TO THE ANALYSIS, THE CON-9437 1451 358 TRIBUTION OF THE EXTERNAL RESONANCES WAS CALCULATED BY USING THE 9437 1451 359 SET OF THE 0 KEV TO 1 KEV KNOWN RESONANCES, SHIFTED IN THE ENERGY 9437 1451 360 RANGES -1 KEV TO 0 KEV, 2 KEV TO 3KEV AND 3 KEV TO 4 KEV; EQUIVA- 9437 1451 361 LENT CONTRIBUTION WAS OBTAINED BY USING 3 FICTICIOUS RESONANCES 9437 1451 362 BELOW 1 KEV AND 3 FICTICIOUS RESONANCES ABOVE 2 KEV(SEE DETAILS IN9437 1451 363 REF./13/). THE ANALYSIS WAS PERFORMED ON THE THICK AND MEDIUM 9437 1451 364 SAMPLE TRANSMISSIONS OF HARVEY DATA (THE THIN SAMPLE DATA WAS NOT 9437 1451 365 USEFUL IN THE HIGH ENERGY RANGE) AND ON THE 1988 FISSION DATA RE- 9437 1451 366 LEASED BY WESTON AT THE BEGINNING OF 1991/7/. THE DEFINITIVE 9437 1451 367 SAMMY FITS WERE PERFORMED IN APRIL 1992 AFTER RENORMALIZATION OF 9437 1451 368 THE 1988 DATA OF WESTON ON THE ENDF/B-VI STANDARD VALUES BETWEEN 19437 1451 369 KEV AND 2 KEV, IN AGREEMENT WITH THE 1991 NEW MEASUREMENTS OF 9437 1451 370 WESTON AND TODD. 9437 1451 371 THE AVERAGE CROSS SECTIONS CALCULATED FROM THE RESONANCE 9437 1451 372 PARAMETERS ARE COMPARED TO THE EXPERIMENTAL VALUES IN THE FOLLO- 9437 1451 373 WING TABLE: 9437 1451 374 9437 1451 375 9437 1451 376 ************************************************************** 9437 1451 377 CROSS-SECTIONS(BARN) 9437 1451 378 --------------------------------------------------- 9437 1451 379 TOTAL FISSION CAPTURE 9437 1451 380 ENERGY -------------- ------------- ------------- 9437 1451 381 KEV CALC(A) EXP(B) CALC(A) EXP(C) CALC(A) EXP(D) 9437 1451 382 ************************************************************** 9437 1451 383 1.0-1.1 24.47 24.95 5.549 5.581 4.728 5.04 9437 1451 384 1.1-1.2 22.82 23.10 5.985 6.017 3.757 2.95 9437 1451 385 1.2-1.3 22.29 22.90 4.601 4.501 4.287 4.00 9437 1451 386 1.3-1.4 22.63 22.85 6.997 6.997 3.012 2.52 9437 1451 387 1.4-1.5 20.42 20.95 4.041 4.059 3.450 3.57 9437 1451 388 1.5-1.6 18.30 18.95 2.564 2.613 3.521 3.89 9437 1451 389 1.6-1.7 21.82 21.90 3.952 3.955 3.833 4.36 9437 1451 390 1.7-1.8 21.26 21.35 3.400 3.425 4.091 4.37 9437 1451 391 1.8-1.9 23.76 23.30 5.178 5.187 3.639 3.14 9437 1451 392 1.9-2.0 18.48 18.90 2.152 2.180 3.205 4.06 9437 1451 393 ************************************************************** 9437 1451 394 1.0-2.0 21.63 21.92 4.442 4.446 3.752 3.79 9437 1451 395 ************************************************************** 9437 1451 396 (A) TOTAL, FISSION AND CAPTURE CROSS SECTIONS CALCULATED BY 9437 1451 397 RESENDD FROM THE RESONANCE PARAMETERS. 9437 1451 398 (B) EXPERIMENTAL TOTAL CROSS SECTIONS FROM REFERENCE/19/. 9437 1451 399 (C) WESTON AND TODD 1988 HIGH RESOLUTION FISSION CROSS SECTIONS 9437 1451 400 FROM REFERENCE/7/ NORMALIZED TO ENDF/B-VI STANDARD IN THE 9437 1451 401 ENERGY RANGE FROM 1.0 KEV TO 2.0 KEV. 9437 1451 402 (D) GWIN 1971 EXPERIMENTAL DATA NORMALIZED TO GWIN 1976 DATA. 9437 1451 403 ************************************************************** 9437 1451 404 9437 1451 405 THE DIFFERENCE OF 1.3% BETWEEN THE AVERAGE CALCULATED TOTAL 9437 1451 406 CROSS SECTION AND THE AVERAGE EXPERIMENTAL CROSS SECTION IN THE 9437 1451 407 ENERGY RANGE FROM 1.0 KEV AND 2.0 KEV IS MAINLY DUE TO THE METHOD 9437 1451 408 OF EVALUATING THE TOTAL CROSS SECTION FROM THE EFFECTIVE CROSS 9437 1451 409 SECTION IN REFERENCE/19/. THE ACCURACY OF THE SAMMY FIT OF THE 9437 1451 410 EXPERIMENTAL TRANSMISSION DATA IS BETTER THAN 0.5% ON THE CROSS 9437 1451 411 SECTION. THE CALCULATED FISSION CROSS SECTIONS ARE IN VERY GOOD 9437 1451 412 AGREEMENT WITH THE EXPERIMENTAL DATA. THE CAPTURE DATA /1/ ARE 9437 1451 413 AVERAGE VALUES OBTAINED FROM THE DATA AVAILABLE IN THE EXFOR FILE 9437 1451 414 AND NORMALIZED TO GWIN 1976 AVERAGE VALUES; THERE ARE LARGE 9437 1451 415 DIFFERENCES BETWEEN THE CALCULATED DATA AND THE EXPERIMENTAL DATA 9437 1451 416 AVERAGED OVER 0.1 KEV INTERVALS; BUT ON THE INTERVAL FROM 1.0 KEV 9437 1451 417 TO 2.0 KEV THE AVERAGE VALUES ARE CONSISTENT WITHIN 1.0%. 9437 1451 418 9437 1451 419 ***************************************************************** 9437 1451 420 COMMENTS ON THE 2.0 KEV TO 2.5 KEV REGION 9437 1451 421 ***************************************************************** 9437 1451 422 9437 1451 423 THIS ENERGY RANGE WAS ALSO ANALYSED AT JAERI /13/. NO 9437 1451 424 PRELIMINARY SET OF RESONANCE PARAMETERS WAS AVAILABLE PRIOR TO THE9437 1451 425 ANALYSIS. MORE THAN 90% OF THE RESONANCES, COMPARED TO THE LOW 9437 1451 426 ENERGY RANGE, COULD STILL BE IDENTIFIED IN THE TRANSMISSION DATA 9437 1451 427 BETWEEN 2 KEV AND 2.5 KEV. THEREFORE THE CORRELATED SAMMY ANALYSIS9437 1451 428 OF HARVEY TRANSMISSIONS AND WESTON FISSION WAS STILL FEASIBLE IN 9437 1451 429 THIS ENERGY RANGE. THE RESONANCE PARAMETERS OBTAINED ARE 9437 1451 430 CONSISTENT AND HAS NEARLY THE SAME STATISTICAL PROPERTIES AS THOSE9437 1451 431 OF THE RESONANCES IN THE 0 TO 2 KEV ENERGY RANGE. A QUITE GOOD FIT9437 1451 432 OF THE TRANSMISSION AND FISSION DATA WAS OBTAINED WITHOUT 9437 1451 433 BACKGROUND AND NORMALISATION ADJUSTMENT. HOWEVER, THE CALCULATED 9437 1451 434 FISSION CROSS SECTIONS ARE, ON AVERAGE, 1.4% LOWER THAN THE 9437 1451 435 EXPERIMENTAL VALUES. THIS DIFFERENCE,WHICH HOWEVER IS NOT LARGER 9437 1451 436 THAN THE SYSTEMATIC ERRORS ON THE EXPERIMENTAL DATA, COULD BE DUE 9437 1451 437 TO THE DIFFICULTIES OF IDENTIFYING THE WIDE J=0+ RESONANCES IN THE9437 1451 438 EXPERIMENTAL DATA, BECAUSE THE EFFECTS OF THE INCREASING 9437 1451 439 RESOLUTION AND DOPPLER WIDTHS. PRIOR TO THE SAMMY FITS, THE 9437 1451 440 FISSION DATA OF WESTON AND TODD (1988 HIGH RESOLUTION DATA) WERE 9437 1451 441 NORMALIZED TO THE ENDF/B-VI STANDARD IN THE ENERGY RANGE FROM 1 9437 1451 442 KEV TO 2 KEV. 9437 1451 443 THE CROSS SECTIONS,CALCULATED FROM THE RESONANCE PARAMETERS 9437 1451 444 AND AVERAGED OVER 0.1 KEV INTERVALS,ARE GIVEN IN THE FOLLOWING 9437 1451 445 TABLE: 9437 1451 446 9437 1451 447 ************************************************************** 9437 1451 448 CR0SS-SECTIONS(BARN) 9437 1451 449 ------------------------------------------ 9437 1451 450 TOTAL FISSION CAPTURE 9437 1451 451 ENERGY -------------- -------------- ------- 9437 1451 452 (KEV) CALC(A) EXP(B) CALC(A) EXP(C) CALC(A) 9437 1451 453 ************************************************************** 9437 1451 454 2.0-2.1 17.34 17.30 2.034 2.062 3.223 9437 1451 455 2.1-2.2 20.27 19.80 2.949 2.999 4.051 9437 1451 456 2.2-2.3 19.34 19.10 2.357 2.393 3.324 9437 1451 457 2.3-2.4 21.28 21.20 3.646 3.679 3.640 9437 1451 458 2.4-2.5 20.03 20.60 3.956 4.024 3.128 9437 1451 459 ************************************************************** 9437 1451 460 2.0-2.5 19.65 19.60 2.989 3.031 3.473 9437 1451 461 ************************************************************** 9437 1451 462 (A) TOTAL, FISSION AND CAPTURE CROSS SECTIONS CALCULATED BY 9437 1451 463 RESENDD, 1% ACCURACY, AT 300 K,FROM THE RESONANCE PARAME- 9437 1451 464 TERS. 9437 1451 465 (B) AVERAGE TOTAL CROSS SECTIONS OBTAINED FROM THE AVERAGE 9437 1451 466 EXPERIMENTAL EFFECTIVE CROSS SECTIONS IN REFERENCE/19/. 9437 1451 467 (C) 1988 HIGH RESOLUTION DATA OF WESTON AND TODD NORMALIZED 9437 1451 468 TO ENDF/B-VI STANDARD IN THE ENERGY RANGE FROM 1 KEV TO 9437 1451 469 2 KEV. 9437 1451 470 ***************************************************************** 9437 1451 471 9437 1451 472 9437 1451 473 ***************************************************************** 9437 1451 474 FISSION AND CAPTURE RESONANCE INTEGRALS 9437 1451 475 ***************************************************************** 9437 1451 476 9437 1451 477 THE FISSION AND CAPTURE RESONANCE INTEGRALS ARE COMPARED TO 9437 1451 478 JENDL-3 DATA IN THE FOLLOWING TABLE: 9437 1451 479 9437 1451 480 ********************************************************** 9437 1451 481 ENERGY RANGE(EV) FISSION(BARN) CAPTURE(BARN) 9437 1451 482 ********************************************************** 9437 1451 483 JENDL-3 PRESENT JENDL-3 PRESENT 9437 1451 484 0.5 - 5.0 85.725 84.879 28.651 28.723 9437 1451 485 5.0 - 10.0 25.081 25.147 19.059 18.950 9437 1451 486 10.0 - 50.0 96.856 99.715 77.181 74.686 9437 1451 487 50.0 - 100.0 40.479 41.552 25.930 25.376 9437 1451 488 100.0 - 301.0 19.677 20.252 17.952 17.729 9437 1451 489 301.0 -1000.0 10.047 10.317 8.348 8.418 9437 1451 490 1000.0 -2000.0 3.484 3.206 2.840 2.634 9437 1451 491 2000.0 -2.E+07 17.783 (17.783) 5.224 (5.224) 9437 1451 492 ********************************************************** 9437 1451 493 TOTAL 299.132 302.851 185.185 181.739 9437 1451 494 ********************************************************** 9437 1451 495 9437 1451 496 THE JENDL-3 RESONANCE PARAMETERS ARE THOSE OBTAINED IN 1987 IN 9437 1451 497 THE ENERGY RANGE 0 KEV TO 1 KEV. THEY ARE SLIGTHLY DIFFERENT FROM 9437 1451 498 THOSE PUBLISHED IN 1989. THAT EXPLAINS THE SMALL DIFFERENCES OB- 9437 1451 499 SERVED BETWEEN JENDL-3 AND THE PRESENT RESULTS IN THIS ENERGY RAN-9437 1451 500 GE. IN THE ENERGY RANGE 1 KEV TO 2 KEV JENDL-3 IS UNRESOLVED 9437 1451 501 RANGE. THE FISSION AND CAPTURE RESONANCE INTEGRALS CALCULATED 9437 1451 502 FROM ENDF/B-V AND THOSE FOUND IN BNL-325 ARE THE FOLLOWING: 9437 1451 503 9437 1451 504 ENDF/B-V FISSION: 302.13 B CAPTURE: 194.10 B 9437 1451 505 BNL-325 FISSION: 310+-10 B CAPTURE: 200+-20 B 9437 1451 506 9437 1451 507 THE CONSEQUENCE OF CHANGING FROM THE OLD SETS OF RESONANCE 9437 1451 508 PARAMETERS(ENDF/B-V AND PREVIOUS SETS) TO THE NEW SET IS THAT 9437 1451 509 THE CAPTURE RESONANCE INTEGRAL WILL DECREASE BY 6.7% COMPARED 9437 1451 510 WITH ENDF/B-V VALUE. 9437 1451 511 9437 1451 512 REFERENCES OF APPENDIX 9437 1451 513 1) R.GWIN ET AL.,NUCL.SCI.ENG.,45,25(1971) 9437 1451 514 2) R.GWIN ET AL.,NUCL.SCI.ENG.,59,79(1976) 9437 1451 515 3) R.GWIN ET AL.,NUCL.SCI.ENG.,61,116(1976) 9437 1451 516 4) R.GWIN ET AL.,NUCL.SCI.ENG.,88,37(1984) 9437 1451 517 5) J.BLONS, NUCL.SCI.ENG.,51,130(1973) 9437 1451 518 6) L.W.WESTON ET AL.,NUCL.SCI.ENG.88,567(1984) 9437 1451 519 7) L.W.WESTON ET AL.,TO BE PUBLISHED(HIGH RESOLUTION 1988 DATA) 9437 1451 520 8) R.R.SPENCER ET AL.,NUCL.SCI.ENG.,96,318(1987) 9437 1451 521 9) J.A.HARVEY ,MITO 1988,PAGE 115 9437 1451 522 10) A.CARLSON ET AL.,PRELIMINARY RESULTS OF THE ENDF/B-6 STANDARD 9437 1451 523 EVALUATION(SEPT 8 1987) 9437 1451 524 11) H.DERRIEN AND G. DE SAUSSURE,ORNL-TM-10986(1988) 9437 1451 525 12) H.DERRIEN ET AL.,NUCL.SCI.ENG.,106,434(1990) 9437 1451 526 13) H.DERRIEN, J.NUCL>SCI.TECHNOL.,30,845(1993). 9437 1451 527 14) N.M.LARSON ET AL.,ORNL/TM-7485,ORNL/TM-9179,ORNL/TM-9719/R1 9437 1451 528 15) L.LEAL AND R.N.HWANG,TRANS.AM.NUC.SOC.,55,340(1987) 9437 1451 529 16) T.NAKAGAWA,RESENDD A JAERI VERSION OF RESEND,JAER=-M 84-192 9437 1451 530 (1984). 9437 1451 531 17) C.WAGEMANS ET AL.,MITO 1988,PAGE 91 9437 1451 532 18) L.W.WESTON,PRIVATE COMMUNICATION(1992) 9437 1451 533 19) H.DERRIEN,J.NUCL.SCI.TECHNOL.,29,794(1992). 9437 1451 534 ********************************************************** 9437 1451 535 9437 1451 536 9437 1451 537 9437 1451 538 9437 1451 539 ********************************************************** 9437 1451 540 ***************** Program LINEAR (VERSION 2002-1) ***************9437 1451 541 For All Data Greater than 1.0000E-10 barns in Absolute Value 9437 1451 542 Data Linearized to Within an Accuracy of .100000000 per-cent 9437 1451 543 ***************** Program RECENT (VERSION 2002-1) ***************9437 1451 544 for All Data Greater than 1.0000E-10 barns in Absolute Value 9437 1451 545 Data Linearized to within an Accuracy of .100000000 per-cent 9437 1451 546 ***************** Program SIGMA1 (VERSION 2002-1) ***************9437 1451 547 Data Doppler Broadened to 300.000000 Kelvin 9437 1451 548 for All Data Greater than 1.0000E-10 barns in Absolute Value 9437 1451 549 Data Linearized to Within an Accuracy pf .100000000 per-cent 9437 1451 550 ***************** Program FIXUP (Version 2002-1) ****************9437 1451 551 Corrected ZA/AWR in All Sections-----------------------------Yes 9437 1451 552 Corrected Thresholds-----------------------------------------Yes 9437 1451 553 Extended Cross Sections to 20 MeV----------------------------No 9437 1451 554 Allow Cross Section Deletion---------------------------------No 9437 1451 555 Allow Cross Section Reconstruction---------------------------No 9437 1451 556 Make All Cross Sections Non-Negative-------------------------Yes 9437 1451 557 Delete Energies Not in Ascending Order-----------------------Yes 9437 1451 558 Deleted Duplicate Points-------------------------------------Yes 9437 1451 559 Check for Ascending MAT/MF/MT Order--------------------------Yes 9437 1451 560 Check for Legal MF/MT Numbers--------------------------------Yes 9437 1451 561 Allow Creation of Missing Sections---------------------------No 9437 1451 562 Allow Insertion of Energy Points-----------------------------No 9437 1451 563 Create Uniform Energy Grid-----------------------------------No 9437 1451 564 Delete Section if Cross Section =0 at All Energies-----------Yes 9437 1451 565 ***************** Program GROUPIE (VERSION 2002-1) **************9437 1451 566 Unshielded Group Averages Using 640 Groups 9437 1451 567 Weighting Spectrum: Flat (Constant) Spectrum 9437 1451 568 1 451 572 19437 1451 569 2 151 4 39437 1451 570 3 18 217 19437 1451 571 33 18 18 09437 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19400000.0 2.24245169 19500000.0 2.244551839437 3 18 215 19600000.0 2.24665196 19700000.0 2.24875209 19800000.0 2.250852229437 3 18 216 19900000.0 2.25295235 20000000.0 0.0 9437 3 18 217 9437 3 099999 9437 0 0 0 9.42390E+4 2.36999E+2 0 0 0 1943733 18 1 0.000000+0 0.000000+0 0 18 0 4943733 18 2 0.000000+0 0.000000+0 0 1 16 8943733 18 3 1.000000-5 0.000000+0 1.000000+0 2.500000-3 3.000000+2 3.600000-3943733 18 4 3.000000+4 4.900000-3 1.000000+5 6.400000-3 2.000000+5 0.000000+0943733 18 5 1.500000+7 1.000000-2 2.000000+7 0.000000+0 943733 18 6 0.000000+0 0.000000+0 0 1 8 4943733 18 7 1.000000-5 0.000000+0 1.000000+0 4.000000-4 1.500000+7 0.000000+0943733 18 8 2.000000+7 0.000000+0 943733 18 9 0.000000+0 0.000000+0 0 1 18 9943733 18 10 1.000000-5 0.000000+0 2.000000+5 4.900000-5 1.000000+6 4.900000-5943733 18 11 2.000000+6 4.900000-5 3.000000+6 1.400000-4 6.000000+6 1.000000-4943733 18 12 8.000000+6 2.250000-4 1.500000+7 0.000000+0 2.000000+7 0.000000+0943733 18 13 0.000000+0 0.000000+0 1 5 21 6943733 18 14 1.000000-5 2.530000-2 9.000000-2 2.500000-1 1.000000+0 2.000000+7943733 18 15 1.000000-4 3.230000-5 3.420000-5 3.230000-5 0.000000+0 3.600000-5943733 18 16 3.390000-5 3.200000-5 0.000000+0 6.400000-5 3.380000-5 0.000000+0943733 18 17 2.560000-4 0.000000+0 0.000000+0 943733 18 18 943733 099999 9437 0 0 0 0 0 0 0 9.52410E+4 2.38986E+2 0 1 34 109543 1451 1 0.0 1.00000E+0 0 0 0 69543 1451 2 1.00000E+0 2.00000E+7 0 0 10 20029543 1451 3 3.00000E+2 0.0 1 0 110 49543 1451 4 95-Am-241 JAERI EVAL-MAR88 T.NAKAGAWA 9543 1451 5 DIST-Feb2004 9543 1451 6 ----IRDF-2002 MATERIAL 9543 9543 1451 7 -----INCIDENT NEUTRON DATA 9543 1451 8 ------ENDF-6 FORMAT 9543 1451 9 **************************************************************** 9543 1451 10 95-AM-241 JAERI EVAL-MAR88 T.NAKAGAWA 9543 1451 11 JAERI-M 89-008 DIST-SEP89 9543 1451 12 ----JENDL-3.2 MATERIAL 9543 9543 1451 13 9543 1451 14 AM-241 FISSION 9543 1451 15 **************************************************************** 9543 1451 16 HISTORY 9543 1451 17 82-03 EVALUATION FOR JENDL-2 WAS MADE BY Y.KIKUCHI (JAERI) /1/. 9543 1451 18 88-03 RE-EVALUATION FOR JENDL-3 WAS MADE BY T.NAKAGAWA (JAERI) 9543 1451 19 /2/. 9543 1451 20 9543 1451 21 MF=1 GENERAL INFORMATION 9543 1451 22 MT=451 COMMENT AND DICTIONARY 9543 1451 23 9543 1451 24 MF=2,MT=151 RESONANCE PARAMETERS 9543 1451 25 RESOLVED RESONANCES FOR MLBW FORMULA : 1.0E-5 - 150 EV 9543 1451 26 DATA OF DERRIEN AND LUCAS /5/ WERE ADOPTED AND 5 NEGATIVE 9543 1451 27 RESONANCES WERE ADDED. VALUES OF TOTAL SPIN J WERE 9543 1451 28 REPLACED WITH ARBITRARILY ASSUMED VALUES. 9543 1451 29 9543 1451 30 UNRESOLVED RESONANCES : 150 EV - 30 KEV 9543 1451 31 PARAMETERS WERE DETERMINED BY USING ASREP/6/ SO AS TO 9543 1451 32 REPRODUCE THE CAPTURE CROSS SECTION MEASURED BY VANPRAET 9543 1451 33 ET AL. /7/ AND THE FISSION CROSS SECTION BY DABBS ET AL. 9543 1451 34 /8/. 9543 1451 35 ENERGY INDEPENDENT PARAMETERS: 9543 1451 36 R=9.37 FM, GAM-G=0.044 EV, DOBS=0.4 EV 9543 1451 37 ENERGY DEPENDENT PARAMETERS: 9543 1451 38 AT 150 EV: S0= 1.08E-4, S1=2.72E-4, WF=0.24 MILLI-EV 9543 1451 39 AT 30 KEV: S0= 0.79E-4, S1=1.99E-4, WF=0.30 MILLI-EV 9543 1451 40 9543 1451 41 CALCULATED 2200-M/S CROSS SECTIONS AND RESONANCE INTEGRALS 9543 1451 42 2200 M/S VALUE RES. INT. 9543 1451 43 TOTAL 614.6 B - 9543 1451 44 ELASTIC 11.13 B - 9543 1451 45 FISSION 3.018 B 13.9 B 9543 1451 46 CAPTURE 600.4 B 1300 B 9543 1451 47 9543 1451 48 MF=3 NEUTRON CROSS SECTIONS 9543 1451 49 9543 1451 50 MT=1,2 TOTAL AND ELASTIC SCATTERING CROSS SECTIONS 9543 1451 51 CALCULATED WITH OPTICAL AND STATISTICAL MODELS BY USING 9543 1451 52 CASTHY/9/. OPTICAL POTENTIAL PARAMETERS/10/ WERE OBTAINED9543 1451 53 BY FITTING THE DATA OF PHILLIPS AND HOWE /11/ : 9543 1451 54 V = 43.4 - 0.107*EN (MEV) 9543 1451 55 WS= 6.95 - 0.339*EN + 0.0531*EN**2 (MEV) 9543 1451 56 WV= 0 , VSO = 7.0 (MEV) 9543 1451 57 R = RSO = 1.282 , RS = 1.29 (FM) 9543 1451 58 A = ASO = 0.60 , B = 0.5 (FM) 9543 1451 59 9543 1451 60 9543 1451 61 MT=18 FISSION CROSS SECTION 9543 1451 62 EVALUATED ON THE BASIS OF THE DATA BY DABBS ET AL./8/ 9543 1451 63 9543 1451 64 MF=33 COVARIANCES OF NEUTRON CROSS SECTIONS FROM JENDL/D-99 9543 1451 65 MT=18 9543 1451 66 ADOPTED FROM ENDF/B-VI/12/. 9543 1451 67 9543 1451 68 9543 1451 69 REFERENCES 9543 1451 70 1) KIKUCHI Y.: JAERI-M 82-096 (1982). 9543 1451 71 2) NAKAGAWA T.: JAERI-M 88-008 (1989). 9543 1451 72 5) DERRIEN H. AND LUCAS B.: 1975 WASHINGTON, P.637, 9543 1451 73 NBS-SP-425 (1975). 9543 1451 74 6) KIKUCHI Y.: PRIVATE COMMUNICATION. 9543 1451 75 7) VANPRAET G. ET AL.: 1985 SANTA FE, 1, 493 (1986). 9543 1451 76 8) DABBS J.W.T. ET AL.: NUCL. SCI. ENG., 83, 22 (1983). 9543 1451 77 9) IGARASI S.: J.NUCL.SCI.TECHNOL.,12,67 (1975). 9543 1451 78 10) IGARASI S. AND NAKAGAWA T.: JAERI-M 8342 (1979). 9543 1451 79 11) PHILLIPS T.W. AND HOWE R.E.: NUCL. SCI. ENG., 69, 375 (1979). 9543 1451 80 12) ZHOU DELIN, GU FUHUA ET AL.: ENDF/B-VI (MAT=9543) (1988). 9543 1451 81 **************************************************************** 9543 1451 82 9543 1451 83 9543 1451 84 9543 1451 85 9543 1451 86 ***************** Program LINEAR (VERSION 2002-1) ***************9543 1451 87 For All Data Greater than 1.0000E-10 barns in Absolute Value 9543 1451 88 Data Linearized to Within an Accuracy of .100000000 per-cent 9543 1451 89 ***************** Program RECENT (VERSION 2002-1) ***************9543 1451 90 for All Data Greater than 1.0000E-10 barns in Absolute Value 9543 1451 91 Data Linearized to within an Accuracy of .100000000 per-cent 9543 1451 92 ***************** Program SIGMA1 (VERSION 2002-1) ***************9543 1451 93 Data Doppler Broadened to 300.000000 Kelvin 9543 1451 94 for All Data Greater than 1.0000E-10 barns in Absolute Value 9543 1451 95 Data Linearized to Within an Accuracy pf .100000000 per-cent 9543 1451 96 ***************** Program FIXUP (Version 2002-1) ****************9543 1451 97 Corrected ZA/AWR in All Sections-----------------------------Yes 9543 1451 98 Corrected Thresholds-----------------------------------------Yes 9543 1451 99 Extended Cross Sections to 20 MeV----------------------------No 9543 1451 100 Allow Cross Section Deletion---------------------------------No 9543 1451 101 Allow Cross Section Reconstruction---------------------------No 9543 1451 102 Make All Cross Sections Non-Negative-------------------------Yes 9543 1451 103 Delete Energies Not in Ascending Order-----------------------Yes 9543 1451 104 Deleted Duplicate Points-------------------------------------Yes 9543 1451 105 Check for Ascending MAT/MF/MT Order--------------------------Yes 9543 1451 106 Check for Legal MF/MT Numbers--------------------------------Yes 9543 1451 107 Allow Creation of Missing Sections---------------------------No 9543 1451 108 Allow Insertion of Energy Points-----------------------------No 9543 1451 109 Create Uniform Energy Grid-----------------------------------No 9543 1451 110 Delete Section if Cross Section =0 at All Energies-----------Yes 9543 1451 111 ***************** Program GROUPIE (VERSION 2002-1) **************9543 1451 112 Unshielded Group Averages Using 640 Groups 9543 1451 113 Weighting Spectrum: Flat (Constant) Spectrum 9543 1451 114 1 451 118 19543 1451 115 2 151 4 19543 1451 116 3 18 217 19543 1451 117 33 18 16 09543 1451 118 9543 1 099999 9543 0 0 0 9.52410E+4 2.38986E+2 0 0 1 09543 2151 1 9.52410E+4 1.00000E+0 0 0 1 09543 2151 2 1.00000E-5 1.50000E+2 0 0 0 09543 2151 3 2.50000E+0 9.37000E-1 0 0 0 09543 2151 4 9543 2 099999 9543 0 0 0 9.52410E+4 2.38986E+2 0 0 0 09543 3 18 1 1.99999E+8 1.99999E+8 0 0 1 6419543 3 18 2 641 1 9543 3 18 3 .000100000 51.3052773 .000105000 50.1079487 .000110000 48.97045709543 3 18 4 .000115000 47.9222736 .000120000 46.6968760 .000127500 45.33727459543 3 18 5 .000135000 44.0950790 .000142500 42.9534963 .000150000 41.71795199543 3 18 6 .000160000 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